RS-16-248, Byron/Braidwood Nuclear Stations, Revision 16 to Updated Final Safety Analysis Report, Appendix a - Application of NRC Regulatory Guides
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B/B-UFSAR A1.1-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.1 NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS
The Licensee meets a ll objectives set fo rth in Revision 0 of Regulatory Guide 1.1 (Safety Gui de 1) as presented in Subsection 6.3.2.2.
B/B-UFSAR A1.2-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.2 THERMAL SHOCK TO REACTOR PRESSURE VESSELS
Westinghouse follows all recom mendations of Revision 0 of Regulatory Guide 1.2 (Sa fety Guide 2). The gu ide Position C.1 is followed by Westinghou se's own analytical and experimental programs as well as by p articipation in the He avy Section Steel Technology (HSST) Program at Oak Ridge National Laboratory.
Analytical techniques ha ve been developed by Westinghouse to perform fracture evalu ations of reactor vessels under thermal shock loadings.
Under the heavy section steel technology progr am, a number of 6-inch thick 39-inch OD steel pressure vessels containing carefully prepared and sharpened surface cracks are being tested.
Test conditions include both hydraulic inter nal pressure loadings and thermal shock loadings. The obj ective of this program is to validate analytical frac ture mechanics techniques and demonstrate quantitatively the margin of safety inherent in reactor pressure vessels. A number of vessels have been tested under hydraulic pressure loadings, and results have confi rmed the validity of fracture analysis techniques. The resu lts and implications of the
hydraulic pressure tests are s ummarized in Oak Ridge National Laboratory report ORNL-TM-T5909.
Six thermal shock expe riments have been comp leted and are now being evaluated. For representative conditi ons, flaws are shown to initiate and arrest in a predictable manner.
These tests have demonstrated the applicability of presently used fracture assessment procedures to both high and l ow toughness vessel, as shown in reports ORNL/NU REG-40 and ORNL-6187.
Fracture toughness testing of ir radiated compact tension fracture toughness specimens has been completed.
The complete postirradiation data on 0.394-inch, 2-in ch, and 4-inch thick specimens are now available from the HSST program. Both static and dynamic postirradiation frac ture toughness data have been obtained. Evaluation of the data obtained to date on material irradiated to fluences b etween 2.2 a nd 4.5 x 10 19 n/cm 2 indicates that the reference toughness curve as contained in the ASME Section III Code remains a conse rvative lower bound for toughness values for pressure vessel steels.
Details of progress and results obtained in the HSST program are available in the hea vy section steel tec hnology program progress reports, issued by Oak Ridge Nat ional Laboratory.
B/B-UFSAR A1.2-2 REVISION 4 - DECEMBER 1992 Regulatory Position C.2 is followed inasmuch as no significant changes have been made in approved core or reactor designs.
The guide position C.3 is follow ed since the ves sel design does not include the use of an engineerin g solution to assure adequate recovery of the fracture toughne ss properties of the vessel material. If additional margin is needed, the r eactor vessel can be annealed at a ny point in its service life. This solution is already feasible, in p rinciple, and could be performed with the vessel in place.
An assessment of pressurized thermal shock events for the Byron and Braidwood Statio ns has been docu mented in response to 10 CFR 50.61, which app ears in Subsecti on 5.3.1.5.1.
These requirements remain in e ffect even though Regulatory Guide 1.2 was withdrawn on June 17, 19
- 91. It has been superseded by 10 CFR 50.61 and Regula tory Guide 1.154.
B/B-UFSAR A1.3-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.3 ASSUMPTIONS USED FOR E VALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR BOILING WATER REACTORS
This guide is pertin ent to BWRs only.
B/B-UFSAR A1.4-1 REVISION 12 - DECEMBER 2008 REGULATORY GUIDE 1.4 ASSUMPTIONS USED FOR E VALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR PRESSUR IZED WATER REACTORS
The requirements in Revision 2 of this guide h ave been adhered to in all pertinent sections of this applic ation. The meteorology assumptions from the gui de are detailed in S ubsections 2.3.4 and 2.3.5. The guide assumptions on radioisotope releases are detailed in Section 15.6
.5, as are the assumpt ions on containment spray effectiveness.
This guide, although u sed in the original plant design, has been superseded by Regulatory Guide 1
.183, "Alternative Radiological Source Terms for Evalu ating Design Basis Acc idents at Nuclear Power Reactors".
B/B-UFSAR A1.5-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.5 ASSUMPTIONS USED FOR E VALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A STEAM LINE BREAK ACCIDENT FOR BOILING WATER REACTORS
Regulatory Guide 1.5 (Safety Gui de 5) is pertinent to BWRs only.
B/B-UFSAR A1.6-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.6 INDEPENDENCE BETWEEN RED UNDANT STANDBY (ONSITE) POWER SOURCES AND BETWEEN THEIR DISTRIBUTION SYSTEMS The Licensee complies with Revision 0 of thi s regulatory guide.
Refer to Subsections 8.1.1 and 8.1.6 for further information.
B/B-UFSAR A1.7-1 REVISION 11 - DECEMBER 2006 REGULATORY GUIDE 1.7 CONTROL OF COMBUSTIBLE G AS CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS-OF-COOLANT ACCIDENT Per 10 CFR 50.44 and T echnical Specification Amendments Nos. 143 and 137 for Byron Statio n, Units 1 and 2 and Braidwood Station, Units 1 and 2, respectiv ely, Regulatory Guide 1.7 Revision 2 is no longer applicable for Byron and Braidwood. See Subsection 6.2.5 for further discussion.
B/B-UFSAR A1.8-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.8 PERSONNEL SELECTION AND TRAINING
The Licensee complies with Revision 1 of thi s regulatory guide.
Refer to Sections 13.1 and 13.2 for further information. In addition, the STA training program incorporates Revision 2 of this regulatory guide, which has more restri ctive requirements than Revision 1.
Training and personnel selection for the rad iation protection program is discussed in Section 12.1 and Section 12.5.
B/B-UFSAR A1.9-1 REVISION 15
- DECEMBER 2014 REGULATORY GUIDE 1.9 SELECTION, DESIGN, QUA LIFICATION AND TESTING OF DIESEL-GENERATOR UNITS USED AS CLASS 1E ONSITE ELECTRIC POWER SYSTEMS AT NUCLEAR POWER PLANTS
Regulatory Guide (RG) 1.9, Rev ision 3, endorses IEEE Standard 387-1984, "IEEE Standard Criteria for Diesel G enerator Units Applied as Standby P ower Supplies for Nu clear Power Generating Stations." In addition to this standard, RG 1.9, Revision 3, provides supplemental regulatory positions. The Licensee complies with IEEE S tandard 387-1984 a nd these supplemental regulatory positions in Revision 3 w ith the following clarifications regarding:
- 1. Regulatory Position C.1.4 Due to the high transfor mer in-rush current, the voltage may dip belo w the required limit of 75% of nominal upon ene rgizing the 480-Volt substation transformers and the ir auxiliary loads.
However, this dip is of a very short durat ion (0.2 to 0.5 seconds) and will occur immediately a fter the diesel generator breaker is closed. Since there is a 1
.8-to-2.0 second delay between diesel generator breaker closure and the reset of the bus undervoltage relays, the voltage will
have recovered to the required limits prior to the beginning of load sequencing.
Exception is taken to the minimum frequency design requirement during e mergency load se quencing of 95 percent of nominal (57 Hz). The des ign limitations of available emergency diesel g enerator (EDG) replacement subcomponents, in conjun ction with e xisting EDG performance limitations, may preclude ac hieving this transient loading performanc e requirement as a result of maintenance or modifi cation. This situation has manifested itself at Byron Station as a result of the modification of the obsolete EDG ele ctronic governing system with its mode rnized replaceme nt. Subsequent modification testing and model ing determined that the performance limitations of the new electronic governor, in conjuncti on with combustion air (turbocharger) limit ations, resulted in a momentary frequency drop below 57 Hz during the s tart of the electric-driven auxiliary fe edwater pump. This frequency drop could not be compensated for by the
governor dynamic adjustment or equipment maintenance.
Evaluations have bee n performed that justify short duration frequency drops below 57 Hz during large pump motor starts. Consideration was given to the effects of the transient on EC CS flow requirements, B/B-UFSAR A1.9-2 REVISION 10
- DECEMBER 2004 pumps/motors, motor-op erated valves, b attery chargers, instrument inverters and die sel generator protection.
It was determined that t he affected systems and components will meet the ir intended safety-related design basis functio ns for short perio ds below 57 Hz during emergency loa d sequencing.
A clarification is provided for the line item, "Frequency should be restore d to within 2 percent of nominal in less than 60 percent of e ach load-sequence interval for step load increase." Starting the containment spr ay (CS) pump in the load sequence is dependent on when the Hi-3 containment pressure signal is r eached. As descri bed in Table 8.3-5, for train A, if containment spr ay actuation is not required at 27 seconds after a safety injection actuation signal, automatic start of the CS pump is blocked until all ot her loads are sequ enced on to the EDG. Train A has an add itional auxiliary feedwater (AF) pump load. Therefo re, the CS pump starting time in the load sequence may either be at 27 seconds, which is the more severe EDG load case, or after 52 seconds. At the 27-second sequence case, the frequency is restored as recommended in the Regulatory Guide. In the greater than 52-second s equence case, there is a possibility of a 5-second load-sequence interval between the AF pump start and the CS pump start. EDG performance modeling under full-flow conditions indicates that the AF pump exceeds the 60-percent loading interval frequency r ecovery guidelin es. Even though the frequ ency undershoot has cleared, a potential frequency overshoo t greater than 2 percent of nominal frequency may still be mome ntarily present beyond the 60-percent lo ading interval frequency guideline. Although a time interval of greater than 60 percent of the loading in terval may be required for the frequency to recover, worst-case analytical analysis and actual EDG testing have d emonstrated that the frequency will r ecover to within 2 percent of nominal prior to the potenti al loading of the CS pump on to the EDG. Once the load sequencing has been completed, the EDG will operate at n ominal frequency
+/-2 percent, as described in the Regulatory Guide.
- 2. Regulatory Position C.2.2 Exception is taken to the last sentence in this paragraph, "Jumpers and other nonstandard configurations or ar rangements should not be used subsequent to initial equipm ent startup cond itions."
In order to successfully accomplish certain diesel tests it is necessary to use jumpers to simulate B/B-UFSAR A1.9-2a REVISION 7 - DECEMBER 1998 particular engine signals.
The use of jumpers is a normal practice in d iesel engine testing, and the safe use of the jumpers is ensured with d etailed procedures, B/B-UFSAR A1.9-3 REVISION 6 - DECEMBER 1996 which include independent ve rification of circuit restoration.
- 3. Regulatory Position C.2.2.1, "Start Test" Each EDG undergoes a startup test on a monthly basis from "standby conditions." Once every 6 months this test is supplemented by verifying proper start up from "normal standby condit ions." This test is covered in paragraph C.2.2.3, and is further discussed in Item 4 below. 4. Regulatory Positions C.2.2.1, C.2.2.
4, C.2.2.5, and C.2.2.6 Exception has been taken against use of the term "standby conditions" to denote "normal standby conditions."
The term "standby condition" is interpreted as any conditional state of the EDG in which the EDG is considered operable. Mo re specifically, standby conditions for an EDG refer to a condi tion whereby the diesel engine lube oil is being continuously circulated and e ngine jacket wat er and lube oil temperatures are consist ent with manufacturer's recommended operatin g range (low lube oil and jacket water temperature al arm settings to the high lube oil and jacket water temperature alarm settings).
The term "normal standby condition" defines a conditional state of the EDG in which lu be oil and jacket water temperatures are within the prescribed temperature bands of these s ubsystems when the EDG has been at rest for an extended period of time with the prelube oil and jacket water circulating systems operational. It should be n oted that the semiannual fast start test described in paragraph C.2.2.3 is performed from "normal s tandby conditions."
- 5. Regulatory Position C.
2.2.6, "Combin ed SIAS and LOOP Tests" Exception is taken to the statement that the EDG be tested for proper response to a LOOP in conjunction with a safety in jection actuation signal (SIAS) in whatever sequence they might occur. The EDGs are tested for respo nse to a loss of off site power (LOOP), to a SIAS, and to a LOOP and SIAS when they occur concurrently. Performing a LOOP/SIAS EDG test in whatever sequence they might occur is beyond the original licensing basis, provides no additional value, and was not included in Regulat ory Guide 1.9, Revision 2 or Regulatory Guide 1.108.
B/B-UFSAR A1.9-4 REVISION 16
- DECEMBER 2016 6. Regulatory Position C.2.2.7, "Single-Load Rejection Test" Exception is taken to the specified power factor requirements of this par agraph. This test is performed with the die sel generator in the isochronous mode of operatio
- n. In this mode of operation, the power fac tor is a function of the generator loads only and cannot be varied by voltage or speed adjustment.
- 7. Regulatory Position C.2.
2.8, "Full-Load Rejection Test" Exception is taken to the power factor range of between 0.8 and 0.9.
In order to ensure the DG is tested under load condit ions that bound design conditions and c omply with the r ecommendations of Regulatory Guide 1.9, testin g will be pe rformed using an upper limit on power factor of 0.89. This power factor range bou nds the actual desig n basis inductive loading the DG would experience. Since this testing is performed with the di esel generator synchronized with offsite power, grid conditions may not permit achieving a power factor of 0.89. In this case, the required power factor limit is not required to be met, however, the power factor shall be mai ntained as close to the limit as practicable.
- 8. Regulatory Position C.
2.3.2.3, "Refueling Outage Testing" Exception is taken to the statement that the overall emergency diesel generator unit design capability should be demonstrated at every refueling outage by performing the tests ide ntified in Table 1 of Regulatory Guide 1.9. R efueling Outage Testing as identified in Table 1 of Regulatory Guide 1.9 is
performed in accordance with the Technical Specifications, and the test interval may be
supplanted with performance-based, risk-informed test intervals. This statement in Regulatory Position C.2.3.2.3 is in accordance with Section 6.5.2 of IEEE Standard 387-1984. By takin g exception to Regulatory Position C.2.3.2.3, exception is also being taken to the statement in Section 6.5
.2 of IEEE Standard 387-1984 that the diesel gen erator unit shall be given one cycle of each of t he specified tests at least once every 18 months to demonstrate its continued capability of performing its required function.
Compliance with the requirements of this guide is described further in Subsections 8.1.2, 8.1.20, 8.3.1.1.
2.2 and 8.3.1.2.
Therefore, the L icensee meets the object ives set forth in this regulatory guide.
B/B-UFSAR A1.10-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.10 MECHANICAL (CADWELD) SPL ICES IN REINFORCING BARS OF CATEGORY I CO NCRETE STRUCTURES Subsection B.2.3 of Ap pendix B describes conformance to the regulatory positions in Revision 1 of Regulato ry Guide 1.10. The requirements remain in effect even though the regulatory guide was withdrawn on July 8, 1981. The regulatory positions are now covered by one or mo re national standards.
B/B-UFSAR A1.11-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.11 INSTRUMENT LINES PENETRATING PRIMARY REACTOR CONTAINMENT The Licensee complies with the requirements of Regulatory Guide 1.11 (Safety Guide 1
- 1) and its supplement as discussed in Subsection 7.1.2.5.
B/B-UFSAR A1.12-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.12 INSTRUMENTATION FOR EARTHQUAKES
The seismic instrumentation plans for Byron and Braidwood Stations satisfy the Regulatory Guide 1.12, Re vision 1, requirements for plants with maximum fou ndation accelerations of 0.3g or less. Byron and Braidwood Sta tions have been designed for a foundation acceleration of 0.2
- g. A compar ison between requirements of Regula tory Guide 1.12 and the Byron/Braidwood seismic instrumentation plan is given in the following.
As required by Regulat ory Guide 1.12, Part C Regulatory Position 1, three triaxial time-history accelerographs - one in the free field, one on the cont ainment building f oundation, and one on the containment shell wall are provided.
Additional triaxial time-history accelerograph s ensors are p rovided on the containment refueling fl oor and at the foundat ion of the river screen house (Byron).
One triaxial peak acce lerograph capable of measuring the absolute peak acceleration in three ortho gonal directions, coinciding with the major axes of the analytical building model, is provided at each of the followin g three locations:
- a. on the accumulat or tank in t he containment building,
- b. on the safety injection piping in the containment building, and
- c. on the essential servi ce return piping in the auxiliary building.
These locations satisfy the requirements of Regulatory Guide 1.12, Part C, Regulatory Position 1, Sections a.
1, a.2, and a.3.
A triaxial response sp ectrum recorder capabl e of measuring both horizontal and v ertical motion and capab le of providing signals for immediate control room indication is provided on the containment building base slab. The locatio n and specification of this recorder is in compliance with S ection b, Part C of Regulatory Guide 1.12.
B/B-UFSAR A1.12-2 Two additional triaxial response spectrum recorders, with the same specifications as a bove are provided on the floor of the counting room in the auxiliary b uilding, and on the operating floor of the containment buildin
- g. These comply with the requirements of Sectio ns c.1 and c.2 of Part C, Regulatory Position 1.
Section c.3 of Part C, Regul atory Position 1 calls for a separate triaxial response spectrum rec order capable of measuring both horizontal motions and the vertical motion to be provided at the foundation of an independent Sei smic Category I structure where the response is differ ent from that of the reactor containment structure.
Except for the Byron river screen house, all the structures are founded on rock, and will have the same founda tion response as the containment stru cture. The tria xial time-history accelerograph sensor p rovided at the fou ndation level of the Byron river screen house is us ed to determine the response spectra for this location, using the playback un it provided in the control room.
The specifications of all the re sponse spectrum reco rders, such as dynamic range, fr equency range, damping, etc., satisfy the requirements of Part C, Regulatory Positions 4 and 5.
In general, the seis mic instrumentation plan for Byron and Braidwood Stations complies with Regulatory Guide 1.12.
Additional information on instrumentation for earthquakes is provided in Subsection 3
.7.4.1 and 3.7.4.2.
B/B-UFSAR A1.13-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.13 SPENT FUEL STORAGE F ACILITY DESIGN BASIS
The plant design conforms with t he requirements in Revision 1 of this guide as presented in S ubsections 9.1.2, 9.1.3, 9.1.4, 6.5.1, and 9.4.5.
B/B-UFSAR A1.14-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.14 REACTOR COOLANT PUMP FLYWHEEL INTEGRITY
The design meets the requirements in Revision 1 of Regulatory Guide 1.14 with the ex ceptions noted below.
The shaft and the bearings supporting the flywheel are capable of withstanding any combination of the n ormal operating lo ads, anticipated transients, the design-basis LOCA, and the safe shutdown earthquake loads.
Since the issuance of Regulatory Guide 1.14, Rev ision 1, the NRC Staff has provided to Westinghouse a copy of Draft 2, Revision 2, of Regulatory Guide 1.14 (via an April 12, 197 6, letter from Robert B. Minogue to C. Eicheldinger). This draft was formulated from industry and conc erned parties' comments.
It is significant that the Draft 2 version incorporate s several of the Westinghouse comments on Revision 1.
Since Draft 2 h as not been formally published as Revision 2 of Regulatory Guide 1.14, the exceptions and clarifications (from the ori ginal Westinghouse comments) are provided in the following:
- a. Post-Spin Inspection Westinghouse has shown in WC AP-8163, "Topical Report on Reactor Coolant Pump Integrity in LOCA," that the flywheel would not f ail at 290% of n ormal speed for a flywheel flaw of 1.15 inches or less in length.
Results for a double ended guill otine break at the pump discharge with full separation of pipe ends assumed, show the maximum ov erspeed was calculated in WCAP-8163 to be about 280% of normal speed for the same postulated break, and an as sumed instantaneous loss of power to the rea ctor coolant pump. In comparison with the over speed presented above, the flywheel could withstand a speed up to 2.3 times greater than the flywheel sp in test speed of 125%
provided that flaws no great er than 1.15 inches are present. If the maximum speed were 125% of normal speed or less, the critical flaw size for failure would exceed 6 inche s in length. Nondestructive tests and critical dim ension examinations are all performed before the spin tests. The inspection methods employed (described in W CAP-8163) provide
B/B-UFSAR A1.14-2 assurance that flaws significantly s maller than the critical flaw size of 1.15 i nches for 290% of normal speed would be detecte
- d. Flaws in t he flywheel will be recorded in the presp in inspection program (see WCAP-8163). Flaw growth att ributable to the spin test (i.e., from a single reversal of str ess, up to speed and back), under the most adverse conditions, is about three orders of magnitude smaller than that nondestructive inspe ction techniques are capable of detecting. For these reasons, Westi nghouse does not perform postspin inspe ctions and believes the prespin test inspections are adequate.
- b. Interference Fit Stresses and Excessive Deformation Much of Revision 1 deals wit h stresses in the flywheel resulting from the int erference fit between the flywheel and the shaft.
Because Westing house's design specifies a light interf erence fit b etween the flywheel and the shaft; at zero speed, the hoop stresses and radial stresses at the flywheel bore are negligible. Centeri ng of the flywheel relative to the shaft is accomplished by means of keys and/or centering devices attached to the shaft, and at normal speed, the flywheel is not in contact with the shaft in the sense int ended by Revision 1. Hence, the definition of "Excessive Defor mation," as defined in Revision 1 of Regulatory Gui de 1.14, is not applicable to the Westinghouse design since the enlargement of the bore and subsequent part ial separation of the flywheel from the shaft does not cause unbalance of the flywheel. Extensive Wes tinghouse experience with reactor coolant pump flywheels installed in this fashion has verified the ade quacy of the design.
Westinghouse's posit ion is that combined primary stress levels, as de fined in Revisio n 0 of Regulatory Guide 14 (C.2), (a) and (c), are both conservative and proven and that no c hanges to these st ress levels are necessary. Westinghouse designs to these stress limits and thus does not have permanent distortion of the flywheel bore at nor mal or spin test conditions.
- c. Section B, Discu ssion of Cross Rolling R atio of 1 to 3 Cross Rolling Ratio - Westinghouse's position is that specification of a cross rol ling ratio is unnecessary since past evaluatio ns have shown th at ASME SA-533-B Class 1 materials produced without this requirement have suitable toughness for typical flywheel applications. Proper material selection and specification of minimum mater ial properties in the transverse direction a dequately ensu re flywheel integrity. An
B/B-UFSAR A1.14-3 REVISION 13
- DECEMBER 2010 attempt to gain isot ropy in the flywheel material by means of cross rolli ng is unnecessar y since adequate margins of safety are pr ovided by bo th flywheel material selection (ASME SA-533-B Class 1) and by specifying minimum yield and tensile levels and toughness test values ta ken in the direction perpendicular to the maximum working dir ection of the material.
- d. Section C, Item 1a: Relative to Vacuum-melting and Degassing Process or t he Electroslag Process Vacuum Treatment -
The requirements for vacuum melting and degassing proces s or the electroslag process are not essential in meeting the balance of the Regulatory Position nor do they, in the mselves, ensure compliance with the overall Regulatory Position. The initial Safety Guide 14 stated that the "fly wheel material should be produced by a proc ess that min imized flaws in the material and improves its fracture toughness properties." This is accompl ished by using SA-533 material including v acuum treatment.
- e. Section C, I tem 2b: Westinghouse interprets this paragraph to mean:
Design Speed Definit ion - Design spe ed should be 125%
of normal speed or the speed to whic h the pump motor might be electrically driven by station turbine generator during anticipated transients, whichever is greater. Normal speed is de fined as the synchronous speed of the a-c drive motor at 60 Hz.
- f. Section C, Item 4b: Inservice Inspect ion of Reactor Coolant Pump Flywheel For reactor coolant pump mot or serial nu mbers 4S88P961 and 1S88P961, in lieu of Regul atory Position c.4.b(1) and c.4.b(2), a qualified in-p lace UT examination over the volume from the inner bo re of the fl ywheel to the circle of one-half the o uter radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheel may be conducted at approximately 10 year intervals coinciding with the Inservice Inspection schedule as required by ASME Section XI.
For all other reactor coolant pump motors, in lieu of Regulatory Position c.4.b(1) and c.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius or a surface examination (MT and/or PT) of exposed surfa ces of the removed flywheel may be conducted at an interval not to exceed 20 years.
B/B-UFSAR A1.14-4 REVISION 9 - DECMEBER 2002 The requirements for exa mination procedures and
acceptance crite ria as described in Regulatory Guide 1.14 will be followed.
This inspection program meets the intent of Regulatory Guide 1.14 in assuring the continued integrity of the reactor coolant pump flywheels.
The flywheel integrity is described in Subsect ions 5.4.1.5.2 and 5.4.1.5.3.
B/B-UFSAR A1.15-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.15 TESTING OF REINFORCING BARS FOR CATEGORY I CONCRETE STRUCTURES Subsection 3.8.3.6, Table 3.8-2, and Section B
.2 of Appendix B describe conformance to the regulatory positio ns in Revision 1 of Regulatory Guide 1.15. The requ irements remain in effect even though the regulatory gu ide was withdrawn on J uly 8, 1981. The regulatory positions are now c overed by one or more national standards.
B/B-UFSAR A1.16-1 REVISION 12
- DECEMBER 2008 REGULATORY GUIDE 1.16 REPORTING OF OPERATING I NFORMATION - APPENDIX A TECHNICAL SPECIFICATIONS The reporting of specific operat ing information described in Regulatory Guide 1.16, Revision 4 is no longer applicable to Byron and Braidwood as a result of the issua nce of Technical Specification Amendment Nos. 142 and 136 for Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2, respectively.
The Licensee complies with the reporting req uirements contained in Title 10 of the Code of Federal Regul ations and stations' Technical Specifications and T echnical Require ments Manual.
B/B-UFSAR A1.17-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.17 PROTECTION OF NUCLEAR POWER PLANTS AGAINST INDUSTRIAL SABOTAGE The Industrial Security Plan for the Byron a nd Braidwood Stations has been provided to the Regulatory St aff on a proprietary basis.
No comparison of the plan and this regulatory guide is provided in this response. C ompliance with t he requirements of Revision 1 of Regulatory Guide 1.17 is pr esented in Section 13.6 and Subsection 9.5.2.2.
The requirements re main in effect even though the regul atory guide was withdrawn on May 21, 1991.
B/B-UFSAR A1.18-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.18 STRUCTURAL ACCEPTANCE TE ST FOR CONCR ETE PRIMARY REACTOR CONTAINMENTS The structural acceptance test conformed to the requirements of the ASME Code Section II I, Division 2/ACI 359-80 Article CC-6000, as stated in Sub section 3.8.1.7.2.2.
These requirements supersede those endorsed in Regu latory Guide 1.1 8, which was withdrawn on July 8, 1981.
B/B-UFSAR A1.19-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.19 NONDESTRUCTIVE EXAMI NATION OF PRIMARY CONTAINMENT LINER WELDS The plant design conforms to t he regulatory positions in Revision 1 of Regulatory Guide 1.19 (Safety Guide
- 19) as described in Sections B.5 and B.6. The req uirements remain in effect even though the regulatory gu ide was withdrawn on J uly 8, 1981. The regulatory positions are now c overed by one or more national standards.
B/B-UFSAR A1.20-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.20 COMPREHENSIVE VIBRATION ASSESSMENT PROGR AM FOR REACTOR INTERNALS DURING PREOPERATIONAL AND INITIAL STARTUP TESTING The requirements in Revision 2 of Regulatory G uide 1.20 are met.
Refer to Subsection 3.9.2.4 for further discussion.
B/B-UFSAR A1.21-1 REVISION 16 - DECEMBER 2016 REGULATORY GUIDE 1.21 MEASURING, EVALUATIN G, AND REPORTING RADIOACTIVITY IN SOLID WASTES AND RELEASES OF RADIOACTIVE MATERIALS IN LIQUID AND GASEOUS EFFLUENTS FROM LIGHT-WATER-COOLED NUCLEAR POWER PLANTS
The Licensee conforms with Revision 1 of this regulatory guide with the following clarifications keyed to paragraph numbers in the regulatory position.
Exception is taken to the biannual reporting requiremen t specified in Reg ulatory Guide 1.21, Revision 1 since the reporting r equirement in 10 CFR 50.36a was revised from biannual to annual in 1996 (ref erence 61 FR 39299).
- 1. Hourly meteorolo gical data are recor ded for all periods throughout the year, and quarterly summaries are
reported. A separate meteor ological data base for periods of batch releases is not provided.
- 10. If multiple sample points are given for detection of radioiodine and readings are below the threshold of detection, the thres hold limits are not summed over the number of sample points to give the total release rate.
- 13. Radiological imp act on man is provid ed for the maximum exposed individual.
Appendix B:
E.1. Total body and si gnificant orga n doses to individual from receiving-water-rela ted pathways are provided for the maximum exposed individual.
E.3. Organ doses to individu als in unres tricted areas from radioactive iodine and radioactive material in particulate form from all expo sure pathways of exposure are provided for the max imum exposed individual.
E.4. Total body doses to ind ividuals and populations from direct radiation from the f acility are incorporated for the maximum exposed individual in 10 CFR 20 ca lculations.
E.5. Total body doses to the population and average doses to individuals in the population from all receiving-water-rela ted pathways are not included.
E.6. Total body doses to the population and average doses to individuals in the population from gaseous effluents to a dista nce of 50 miles from the site are not included.
B/B-UFSAR A1.21-2 REVISION 7 - DECEMBER 1998 14. Sensitivities in Appendixes A and B of this guide may not be practicable.
These releases are measured to the lowest levels consis tent with existing technology.
Appendix B:
D.1. The total quant ity of solid waste in cubic meters and curies is summed for each quarter.
Assurance of measuri ng very low levels of radioactivity is subject to interpretation of readout which may be effected by noise level, calibration, radi ation, background, etc. In addition, the instrument sensi tivity required to assure compliance with the gu ide may not be a vailable with current technology.
B/B-UFSAR A1.22-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.22 PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS The Licensee complies with Revision 0 of Reg ulatory Guide 1.22 (Safety Guide 22). Refer to Subsections 7.1.2.6, 7.2.2, 7.3.2.2, 8.1.3, 8.3.1.2 and 12.3.4.1 for furt her information.
B/B-UFSAR A1.23-1 REVISION 16 - DECEMBER 2016 REGULATORY GUIDE 1.23 ONSITE METEOROLO GICAL PROGRAMS
The Licensee complies with Revision 1 of Reg ulatory Guide 1.23 with the following clarification:
Based on a commitment made to the NRC during t he implementation of Alternative Source Te rms, finer wind spee d categories provided in the latest appropriate regulatory guidance are to be used the next time the dose consequence calculati ons associated with the LOCA, MSLB, CREA, LRA, S GTR, and FHA events are revised. Finer wind speed categories from Regulatory Guide 1.
23 Rev. 1 have been used in the latest r evised analyses.
Refer to Subsection 2.3.3 for further information.
B/B-UFSAR A1.24-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.24 ASSUMPTIONS USED FOR E VALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A PRESSURIZED WATER REACTOR RADIOACTIVE GAS STORAGE TANK FAILURE
The Licensee complies wi th the regulatory po sition in Revision 0 of Regulatory Guide 1.24 (Safety Guide 2
- 4) as presented in Subsection 15.7.1.3.
B/B-UFSAR A1.25-1 REVISION 12 - DECEMBER 2008 REGULATORY GUIDE 1.25 ASSUMPTIONS USED FOR E VALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT IN THE FUEL H ANDLING AND STORAGE FACILITY FOR BOILING AND PRESSURIZED WATER REACTORS The NSSS vendor's practice and r ecommendations are in agreement with Revision 0 of Regulatory Guide 1.
25 (Safety Guide 25), except that footnote C.1.c cannot be met.
This footnote states that the average burnup for the peak assembl y should be 25,000 MWd/ton or less. The lead r od average burnup for the peak assembly is in the 60,000 MWd/ton range for West inghouse fuel. (See Subsections 15.7.
4.2 and 15.7.4.3.)
This guide, although u sed in the original plant design, has been superseded by Regulatory Guide 1
.183, "Alternative Radiological Source Terms for Evalu ating Design Basis Acc idents at Nuclear Power Reactors".
B/B-UFSAR A1.26-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.26 QUALITY GROUP CLASSIFICA TIONS AND STANDARDS FOR WATER-, STEAM-, AND RADI OACTIVE-WASTE-CONTAINING COMPONENTS OF NUCLEAR POWER PLANTS
The Licensee complies with the regulatory positions stated in Revision 3 of this regulatory guide.
However, excepti ons to this regulatory guide may be taken if replacement components, parts, or materials are no long er available as ASME Section III items.
These replacements will be purchased acc ording to the Exelon Generation Company Qua lity Assurance pro gram but may not be certified to the ASME Code. This practice is consistent with the
NRC Staff position defined in Generic Letter 8 9-09. A list of all components, parts, a nd materials replace d according to this practice are included in Table A 1.26-1. Refer to Subsection 3.2.2 and Table 3.2-1 for ad ditional information on the classification of st ructures, components, and systems.
BYRON-UFSAR A1.26-2 REVISION 8 - DECEMBER 2000 TABLE A1.26-1
COMPONENT, PART, OR MATERIALREFERENCESAFETY CAT. QUALITY GRPELECTRICAL
- 1. Control Room Chiller/R. A. FlahiveICn/aCondenser Tubes to D. Wozniak 9/27/89 BRAIDWOOD-UFSAR A1.26-3 REVISION 9 - DECEMBER 2002 TABLE A1.26-1
ASME SECTION III SUBSTIT UTION COMPONENTS AND PARTS
COMPONENT, PART, OR MATERIALREFERENCESAFETY CAT.QUALITY GRPELECTRICAL Emergency Diesel Generator Lube Oil Strainers 1/2DG02MA/B, 1/2DG03MA/B &
1/2DG06MA/B.
Parts Evaluation No.
A-1991-122-0 I C N/A Emergency Diesel Generator Lube Oil Thermostatic Controller Valves 1/2DG5003A/B.
Parts Evaluation No.
A-1992-159-0 I C N/A Emergency Diesel Generator Engine Crankcase Lube Oil Manual Fill/Drain Isolation Valves 1/2DO072A/B.
Parts Evaluation No.
A-1992-227-0 I C N/A B/B-UFSAR A1.27-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.27 ULTIMATE HEAT SINK FOR N UCLEAR POWER PLANTS
The Licensee meets all objectives set forth in Revision 2 of this regulatory guide as presented in Subsections 2.3.1, 2.4.11, 9.2.5.1, 9.2.5.2, 9.2.5.3, and Technical Specifi cation 3.7.9.
B/B-UFSAR A1.28-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.28 QUALITY ASSURANCE PROG RAM REQUIREMENTS (DESIGN AND CONSTRUCTION)
The Licensee complies with the positions of the regulatory guide with the following exception:
Regulatory Guide 1.2 8, Revision 3 requ ires the licensee to ensure the requiremen ts of NQA-1 are met by its suppliers. Suppliers are audited to NQA-1 or ANSI/ASME N45.2-series standar ds. Because of the large quantity of vendors maintained on the ap proved bidders list, the three year frequency of audits, and some non-ASME vendors who are reluctant to revise their QA program, not all suppliers meet NQA-1.
The Requirements of Regu latory Guide 1.28 wi ll be applied to ANSI/ASME NQA-1-1994 vice th e endorsed ANSI/AS ME NQA-1-1983.
Also refer to the Quality As surance Program Topical Report NO-AA-10.
B/B-UFSAR A1.29-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.29 SEISMIC DESIGN CLASSIFICATION
The Licensee complies with the regulatory positions stated in Revision 3 of this regulatory gu ide. Refer to S ubsections 3.2.1 and 3.10.1.2.1 for fur ther information.
B/B-UFSAR A1.30-1 REVISION 15 - DECEMBER 2014 REGULATORY GUIDE 1.30 QUALITY ASSURANCE RE QUIREMENTS FOR THE INSTALLATION, INSPECTI ON, AND TESTING OF INSTRUMENTATION AND EL ECTRIC EQUIPMENT
Regulatory Guide 1.30 endorsed ANSI N45.2.4, Quality Assurance Requirements for the Installatio n, Inspection, a nd Testing of Instrumentation and Electric Equipment. NQA-1-1 994, Subpart 2.4, Installation, Inspection, and Testing Requiremen ts for Power, Instrumentation, and Control Equ ipment at Nuclear Facilities supersedes this commit ment to Regulatory Guide 1.30 and ANSI N45.2.4 as documented in the Quality Ass urance Topical Report (NO-AA-10). Refer to Subsections 3.11.2, 7.
1.2.8, and 8.1.5 for further information.
B/B-UFSAR A1.31-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.31 CONTROL OF FERRI TE CONTENT IN STAINLESS STEEL WELD METAL The Licensee complies wi th the regulatory posi tion described in Regulatory Guide 1.3 1, Revision 3.
The position concerning the control of delta ferrite in the stainless steel welding is discussed in Subsections 5.2.3, 5.3.1.4, and 6.1.1.1.
B/B-UFSAR A1.32-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.32 CRITERIA FOR SAFETY-RELATED ELECTRIC POWER SYSTEMS FOR NUCL EAR POWER PLANTS The Licensee complies wi th the regulatory po sitions in Revision 2 of this guide with t he following excepti ons/clarifications:
Regulatory Position C.1.a.
See position on Regula tory Guide 1.93.
Regulatory Position C.1.d.
See position on Regu latory Guides 1.6 and 1.75.
Regulatory Position C.1.e.
See position on Regula tory Guide 1.75.
Regulatory Position C.1.f.
See position on Regu latory Guide 1.9.
Regulatory Position C.2.a.
See position on Regula tory Guide 1.81.
Regulatory Position C.2.b.
See position on Regula tory Guide 1.93.
B/B-UFSAR A1.33-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.33 QUALITY ASSURANCE PROG RAM REQUIREMENTS (OPERATION)
The Licensee complies with Revision 2 of this regulatory guide with the exception of regulatory pos itions C.2 and C.4. In lieu of specifying individu al audit frequencies, au dits are conducted on a performance-driven frequency not to exceed 24 months. Refer to Topical Report NO-A A-10 for further information on the Quality Assurance Program at the Byron/Braidwood Stations.
In lieu of the 45.2 daughter documents referenced in ANSI N18.7-1976/ANS 3.2, the following sect ions of NQA-1-1994, which incorporates the requirements of NQA 1989 and NQA 1989 into one document, will be utilized per the following matrix:
A NSI N45.2 Daughter Standard A NSI N45.2.1 Subpart 2.1 A NSI N45.2.2 Subpart 2.2 A NSI N45.2.3 Subpart 2.3 A NSI N45.2.5 Subpart 2.5 A NSI N45.2.6 Per Regulatory Guide 1.28 Revision 3 regulatory position C.1, Basic R equirement 2, Supplement 2S-1, and Appendix 2A-1 will be u tilized; however, NQA-1-1994 will be used instead of NQ A-1-1983 A NSI N45.2.8 Subpart 2.8 A NSI N45.2.9 Per Regulatory Guide 1.28 Revision 3 regulatory position C.2, Basic Req uirement 17 and Supplement 17S-1 will be utilized; however, NQA-1-1994 will be used i nstead of NQA 1983. A NSI N45.2.11 NQ A-1-1994 Basic Requirement 3 and Supplement 3S-1 will be utilized. A NSI N45.2.13 NQ A-1-1994 Basic Requirement 7 and Supplement 7S-1 will be utilized.
B/B-UFSAR A1.34-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.34 CONTROL OF ELECTROSL AG WELD PROPERTIES
The Licensee complies wi th the regulatory po sition in Revision 0 of the guide w henever the electroslag we lding process is used for components made of f erritic or austenitic materials. However, electroslag welding is n ot used for equipmen t purchased on the Licensee's specificati ons. (See Subsect ions 5.3.1.4 and 5.4.2.1.1 for furthe r information.)
B/B-UFSAR A1.35-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.35 INSERVICE INSPECTION OF UNGROUTED TENDONS IN PRESTRESSED CONCRETE CONTAINMENT STRUCTURES The requirements in Regu latory Guide 1.35 ar e incorporated into Section XI, IWL, 1992 Edition, 1992 Addenda, as modified by 10 CFR 50.55a(b)
(2)(viii). The Licensee complies with these requirements. Refer to Subsection 3.8.1
.7.3.2 for further information.
B/B-UFSAR A1.35.1-1 REVISION 7 - DECEMBER 1998 REGULATORY GUIDE 1.35.1 DETERMINING PRESTRESSING FOR CES FOR INSPECTION OF PRESTRESSED CONCRETE C ONTAINMENT STRUCTURES The Licensee complies with the July 19 90 edition of this regulatory guide. Refer to Su bsection 3.8.1.7.3
.2 for further information.
B/B-UFSAR A1.36-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.36 NONMETALLIC THER MAL INSULATION FOR AUSTENITIC S TAINLESS STEEL The Licensee complie s with Revision 0 of this guide.
The NSSS vendor practice meets r ecommendations of Regulatory Guide 1.36 but is more stringent in several re spects as discussed below. (Also see pa ragraph 5.2.3.2 for further information.)
The nonmetallic thermal insulation used on the reactor coolant pressure boundary is spe cified to be made of compounded materials which yield low leachable ch loride and/or fluoride concentrations.
The compounded materials in the form of blocks, boards, cloths, tape s, adhesives, cements, e tc., are silicated to provide protection of au stenitic stainless steels against stress corrosion which may re sult from accidental wetting of the insulation by spillage, minor leakage, or other contamination from the environmental atmosphere. Each lot of insulation material is qualified and analyzed to assure that all of the materials provide a compatible combination for the reactor coolant pressure boundary.
The tests for qualific ation specified by the guide (ASTM C692-71 or RDT M12-IT) a llow use of the tested insulation materials if no more than one of the metallic te st samples crack
- s. Westinghouse rejects the tested insul ation material if any of the test samples cracks.
The vendor procedure is more specific th an the procedures suggested by the guide, in t hat the Westinghou se specification requires determination of leacha ble chloride and fluoride ions
from a sample of the i nsulating materials. The procedures in the guide (ASTM D512 and ASTM D117
- 9) do not differ entiate between leachable and unleacha ble halogen ions.
In addition vendor e xperience indicates that only one of the three methods allowed un der ASTM D512 and ASTM D1179 for chloride and fluoride analysi s is sufficiently ac curate for reactor applications. This is t he "referee" method, which is used by Westinghouse. These requirements are de fined in Westinghouse Process Specificatio n PS-83336KA.
B/B-UFSAR A1.37-1 REVISION 15 - DECEMBER 2014 REGULATORY GUIDE 1.37 QUALITY ASSURANCE REQUIR EMENTS FOR C LEANING OF FLUID SYSTEMS AND ASSO CIATED COMPONENTS OF WATER-COOLED NUCL EAR POWER PLANTS
Regulatory Guide 1.37 endorsed ANSI N45.2.1, Quality Assurance Requirements for Cleaning of F luid Systems a nd Associated Components of Water Co oled Nuclear Power Plants. NQA-1-1994, Subpart 2.1, Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components for Nuclear Power Plants supersedes this commit ment to Regulatory Guide 1.37 and ANSI N45.2.1 as documented in the Quality Ass urance Topical Report (NO-AA-10). Refer to Subsections 5.2.3, 5.4.2.1
.1, and 6.1.1.1 for further information.
B/B-UFSAR A1.38-1 REVISION 15 - DECEMBER 2014 REGULATORY GUIDE 1.38 QUALITY ASSURANCE REQUIREMENTS FOR PAC KAGING, SHIPPING, RECEIVING, STORAGE, AND HANDLING OF ITEMS FOR WATER-COOLED NUCLEAR POWER PLANTS
Regulatory Guide 1.38 endorsed ANSI N45.2.2, Quality Assurance Requirements for Packaging, Ship ping, Receiving, Storage, and Handling of items for Water Cooled Nuclear Power Plants. NQA 1994, Subpart 2.2, Quali ty Assurance Require ments for Packaging, Shipping, Receiving, Storage, and Handling of items for Nuclear Power Plants supersedes this commitment to Regulatory Guide 1.38 and ANSI N45.2.2 as documented in the Quality As surance Topical Report (NO-AA-10). Th ese practices are audi ted for compliance in accordance with the Qu ality Assurance Progra m as described in Topical Report NO-AA-10.
B/B-UFSAR A1.39-1 REVISION 15 - DECEMBER 2014 REGULATORY GUIDE 1.39 HOUSEKEEPING REQUIREME NTS FOR WATER-COOLED NUCLEAR POWER PLANTS Regulatory Guide 1.39 endorsed ANSI N45.2.3, Housekeeping Requirements for Water Cooled Nuclear Power Plants. NQA-1-1994, Subpart 2.3, Quality Assurance Requirements for Housekeeping for Nuclear Power Plants supersedes this com mitment to Regulatory Guide 1.39 and ANSI N45.2.3 as documented in the Quality Assurance Topical Report (NO-AA-10).
The Licensee complies wi th Subpart 2.3 of AN SI/ASME NQA-1-1994.
Facility cleanness, care of ma terial and equipment, fire prevention and p rotection, disposal of d ebris, protection of material and control of access are controlle d by the station.
Normal management atte ntion and periodic audits under the Quality Assurance Program provide the desired result at the Byron and Braidwood Stations.
Independent audits by NRC Region III personnel also contribute to the effectiveness of good housekeeping practices.
Refer to Topical Report NO-AA-10 for further information on the Quality Ass urance Program.
Alternative Clarification Byron/Braidwood shal l comply with Subpart 2.3 of ANSI/ASME NQA 1994 except for the following alternative clarification:
Section 2.2, Classificat ion of Cleanness Zones.
Byron/Braidwood does not hav e any areas meeting the description of zone 1 per table listed.
For the purpose of Foreign Material Exclusion, zo ne designations will be determined based on Work Contr ol and Foreign Material Exclusion program requiremen ts and contr ols implemented consistent with best industry practices.
Section 3.1, In lieu of a written reco rd of the entry and exit of all personnel, Pers onnel Accountability f or Zones 1, 2 and 3 will be contro lled as determined by administrative programmatic controls for th e stations Site Access, Locked Doors, Radiation Wor k Permits, Work Co ntrol and Foreign Material Exclusion program.
B/B-UFSAR A1.40-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.40 QUALIFICATION TESTS OF CONTINUOUS-DUTY MOTORS INSTALLED INSIDE THE CONTA INMENT OF WATER-COOLED N UCLEAR POWER PLANTS NSSS Scope It is the Westinghou se position that mot ors inside containment comply with the qualific ation control requirem ents of Criterion III to Appendix B to 10 CFR 50. These r equirements are satisfied by qualification as described in WCAP-8587 and i ts supplement which contains appropria te EQDPs (equipment qualification data packages) for Westinghouse suppl ied continuous duty motors within the containment. The Licensee is in c ompliance with the objectives of Regulatory Guide 1.40, Revision 0.
Non-NSSS Scope
The Licensee complies with the requirements of Regulatory Guide 1.40, Revision 0, wi th the clarificati on to the regulatory position identified and justified below:
Regulatory Position Cl To the extent practi cable, auxiliary equipment that is part of the inst alled motor assembly should also be qualified in accordance with IEEE 334-1971.
Licensee's Position
Comply with regulatory posit ion, in that to the extent practicable, auxilia ry equipment essential to the safety function of the installed motor assembly will be qualified in accordance with IEEE 334-1971.
Justification of Lic ensee's Position
Nonessential auxiliaries hav e no safety function and should be excluded f rom the requirements.
B/B-UFSAR A1.41-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.41 PREOPERATIONAL TESTING OF REDUNDANT ON SITE ELECTRIC POWER SYSTEMS TO VER IFY PROPER LOAD GROUP ASSIGNMENTS The Licensee complies with Revision 0 of thi s regulatory guide.
Refer to Subsection 8.1.8 for further information.
B/B-UFSAR A1.42-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.42 INTERIM LICENSING POLICY ON AS LOW AS PRACTICABLE FOR GASEOUS RADIOIODINE RE LEASES FROM LI GHT WATER-COOLED NUCLEAR POWER REACTORS
This guide was w ithdrawn on March 18, 1976.
Based on the date that the operating license a pplications were docketed, the following documents were used in lieu of Regulatory Guide 1.42:
Appendix I of 10 CFR 50 and Regu latory Guides 1.109, 1.111, and 1.112.
B/B-UFSAR A1.43-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.43 CONTROL OF STAINLESS STEEL WELD CLADDING OF LOW-ALLOY STEEL COMPONENTS The Licensee complies wi th the requirements in R evision 0 of this guide. Refer to Subsection 5.3.
1.4 for further information.
Westinghouse meets t he intent of Regulat ory Guide 1.43 by requiring qualification of any high heat inp ut process, such as the submerged-arc wide-strip welding process and the submerged-arc-6-wire process used on SA-508 Class 2 material, with a performance test as descr ibed in Regulato ry Position 2 of the guide. No qualifi cations are required by the regulatory guide for SA-533 material and equivalent chemist ry for forging grade SA-506 Cla ss 3 material.
B/B-UFSAR A1.44-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.44 CONTROL OF THE USE OF SE NSITIZED STAINLESS STEEL
The Licensee complies with Revision 0 of this guide with the following clarifications keyed to paragraph numbers in the Regulatory Position.
- 1. The Licensee complies in general with the intent of the requirements of this gui de. With regard to fabrication, shipmen t, storage, and co nstruction, the Applicant requests that contaminants be avoided and cleaning solutions be halide free.
- 2. The Licensee com plies with the requirement in that it specified ASME mater ial specifications which require material to be supplied in the solution annealed condition.
- 3. The Licensee does not agree with this requirement.
Specification of solution annealed material is sufficient.
- 4. The Licensee's specifica tions prohibit t he use of materials that have been exposed to sensitizing temperatures in the range of 800
° to 1500°F.
- 5. Same as Item 4.
- 6. The Licensee does not agree with the requirement to perform intergranular co rrosion tests for each welding procedure. Cont rol of the PWR reactor coolant within the limits of the Technical Requirements Manual (TRM) ensures a benign environment (i.e., low oxygen, low fluoride, and chroride content), and the use of welding filler materials with a minimum fer rite number (FN) of 7.5 FN (approximately 7.5% ferrite content) mitigates the concerns for intergranular s tress corrosion cracking (IGSCC).
The position on Regula tory Guide 1.44 is discussed in part in Subsection 5.2.3.4 (Fabrication and Processi ng of Austenitic Stainless Steel) and in Subsection 6.1.1.1.
B/B-UFSAR A1.45-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.45 REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS The Licensee complies with Revision 0 of this guide with the following clarifications
/exceptions keyed to paragraph numbers in the regulatory position.
- 1. Identified leak sources are piped to either the RC drain tank or a misce llaneous drain tank to be utilized for this purpose only. The temperature of selected drain lines is monitored to identify leaks. Tank inventori es are monitored.
Temperature monitoring is more sensitive to small
leaks than flow rate mon itoring specified in the Position.
- 2. Unidentified leak sour ces are monitored to as accurate an equivale nt flow rate as is practicable.
Containment floor drain and reactor cavity flow monitors for unidentified leakage may not always be accurate to w ithin 1 gpm; how ever, sump level monitoring and p ump run time mon itoring are used as alternate means of monitoring floor drain flow. 3. The following leak d etection systems are provided:
Identified Sources
- a. RC drain tank level in dication and temperature indication of selected inlet lines, or
- b. pressurizer relief t ank level indication and temperature indication of selected inlet lines.
Unidentified Sources
- a. containment floor drain and reactor cavity flow, and sump level, b. containment atmosphere p articulate radioactivity monitoring, and
- c. containment gaseous ra dioactivity monitoring.
- d. VCT level and net chargi ng/letdown flow indication provide quantitative ind ication of R CS leakage.
- e. Containment dry-bulb t emperatures and pressure provide indirect indicat ion of leakage to the containment.
B/B-UFSAR A1.45-2 REVISION 9 - DECEMBER 2002 4. Intersystem leakage between primary and secondary plant is monitored via a ir ejector off-gas radiation monitors. Also, pressurizer and makeup tank levels are monitored to yield total reactor coolant leakage.
Refer to subse ction 5.2.5.5 for additional leakage d etection methods.
- 5. Leak detector sensit ivity is as low as practicable. Refer to S ubsection 5.2.5.2 for additional information on containment radiation monitor sensitivity.
- 6. The containment floor drain and reactor cavity flow indications are designed to remain functional after an SSE and are powered by non-ESF buses. The sump l evel indications are seismically qualified and po wered by ESF buses.
Although the containment radiation monitors are not seismically qualified, they are seismically mounted and can be power ed from safety-related buses, via the n on-safety to safety related cross-tie breake rs, if necessary.
- 7. Conversions to common leakage equivalent are supplied to operators wherever possible.
Conversions to a common leakage equivalent are not possible in all cases.
In these cases, the system is intended prima rily for localization or identification of a le ak with no quantitative implications.
Further information on r eactor coolant pressure boundary leak detection can be found in Subsection 5.2.5.
B/B-UFSAR A1.46-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.46 PROTECTION AGAINST PIPE WHIP INSIDE CONTAINMENT
This regulatory guide was withdrawn on M arch 1, 1985 because the July 1981 revision of Standard Review Plan 3.6.2 provides more current information.
The NRC review is included in the Safety Evaluation Reports, NU REG-0876, dated Februa ry 1982, for Byron and NUREG-1002, date d November 1983, for Braidwood.
B/B-UFSAR A1.47-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.47 BYPASSED AND INOPERABLE STAT US INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEMS The Licensee complies wi th the regulatory po sition in Revision 0 of this guide as discussed in Section 7.5 and Subsections 7.1.2.10 and 8.1.9.
B/B-UFSAR A1.48-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.48 DESIGN LIMITS AND LOAD ING COMBINATIONS FOR SEISMIC CATEGORY I FLU ID SYSTEM COMPONENTS Regulatory Guide 1.48 was withdrawn on March 1, 1985 because the July 1981 revision of the Standard Review Plan 3.9.3 provides more current informati on. The NRC review is included in the Safety Evaluation Repo rts, NUREG-0876, dated February 1982, for Byron and NUREG-1002, dated November 198 3, for Braidwood.
B/B-UFSAR A1.49-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.49 POWER LEVELS OF NUCL EAR POWER PLANTS
The Byron/Braidwood de sign meets the r ecommendations of Regulatory Guide 1.49, R evision 1, since the i nitial power level is less than 3800 MW t and analyses and e valuation are made at assumed core power l evels less than the levels in the guide.
B/B-UFSAR A1.50-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.50 CONTROL OF PREHE AT TEMPERATURE FOR WELDING OF LOW-ALLOY STEEL The Licensee complies with Revision 0 of this guide with the following comments and exceptions keyed to p aragraph numbers in the regulatory position.
1.a. Licensee requires preheat temperatures referenced in applicable codes but does not require a maximum interpass temperature.
To date, it has not been found necessary to speci fy a maximum interpass
temperature.
1.b. Welding proced ures are qualifi ed in the preheat temperature range. It is not possible to consistently maintain pr eheat at the minimum temperature during w elding procedure qualification.
- 2. The Licensee does not agree with this position. It is impossible to maintain preheat temperature during fabrication of spool pieces when four or five welds have been made prior to a complete post-weld heat treatment of the spool piece. The only way this co uld be accomplished would be to have intermittent post-w eld heat treatment which in the case of the higher alloy steel, such as 2-1/4 chrome, may be d etrimental.
- 3. Preheat temperature limit is monitor ed, but not interpass temperature.
Westinghouse considers that this guide appli es to ASME Section III Class 1 Components.
The NSSS vendors' practi ce for Class 1 components is in agreement with the requirements of Regul atory Guide 1.50 except for regulatory positions 1 (b) and 2. For class 2 and 3 components, Westinghouse does not apply any of Regul atory Guide 1.50 recommendations.
B/B-UFSAR A1.50-2 In the case of regulatory position 1 (b), the welding procedures are qualified within the preheat tem perature ranges required by Section IX of the AS ME Code. Westingh ouse qualification procedures.
In the case of regulat ory Position 2, the vendor's position described in WCAP-85 77, "The Application of Preheat Temperature After Welding of Pressure Vess el Steels," has been found acceptable by the NRC.
This WCAP establ ishes the guidelines which permit the component manuf acturer to either maintain the preheat until a post-weld heat treatment or al low the preheat to drop to ambient temperature.
In the case of reactor v essel main structural welds, the practice of maintaining p reheat until the interme diate or final post-weld heat treatment h as been followed by the vend or. In either case, the welds have shown high integrity.
The NSSS vendor meets Re gulatory Position 4 in that, for their components, the examinat ion procedures required by Section III and the inservice inspec tion requirements of Section XI are met. (For further information, see Paragraph 5.3.1.4.)
B/B-UFSAR A1.51-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.51 INSERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 NUCLEAR PO WER PLANT COMPONENTS The guidance in Regula tory Guide 1.51 has been incorporated in the 1974 edition of Section XI of the ASME Boiler and Pressure Vessel Code. Later edit ions of the code were used for preservice and inservice inspections of ASME Code Class 2 and 3 components.
This regulatory guide was withdrawn on J uly 5, 1975 because it was no longer needed.
BYRON-UFSAR A1.52-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.52 DESIGN, TESTING AND MAIN TENANCE CRITERIA FOR ENGINEERED-SAFETY-FEATURE ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ADSORPTION UNITS OF LIGHT-WATER-COOLED NUC LEAR POWER PLANTS
The Licensee complies with Revision 2 of the regulatory position with the following c omments and exceptions keyed to paragraph numbers in Section C of the position:
l.e (Deleted) (Note 1)
2.a Entrained water droplets are not considered credible due to significant q uantities of ductwork with elbows. Water droplets, if present, will i mpinge on ducts and drop out of vertical duct risers as t he air enters exhaust plenums. However, t he auxiliary building exhaust system does contain pre filters which can se rve as demisters.
2.d (Deleted) (Note 1) 2.f The auxiliary buildi ng nonaccessible area exhaust filter system consists of t hree built-up filter trains (one standby) with a rated capacity of 62,000 cfm each. The system design flow is 62,730 cfm, which is within 110% of installed filter rated capac ity. For maintenance purposes each train is divided into three banks with each bank sized for seven filters wide an d three high. Each train has a total of 62 HEPA filter elements.
The auxiliary building accessible ar ea exhaust filter system has a rated capacity of 186,000 cfm. The system design flow is 125,490 c fm. This system consists of four built-up filter trains (one standby), and each train is divided into three banks. E ach bank is size d for seven filters wide and thr ee filters high.
Each train has a total of 62 HEPA filter elements.
2.g ALL ESF filter syste ms have local cont rol panel airflow indication. In additi on, the flow rate of each of the stacks is recorded at the local control panel. The airflow rate through the control room em ergency makeup air filter units and the aux iliary building and fuel handling building exhaust ch arcoal booster fans is continuously sensed.
BYRON-UFSAR A1.52-1a REVISION 8 - DECEMBER 2000 The differential pressure ac ross all of the ESF filter unit fans is indicated on lo cal control panels.
High and low differential press ure, and fan trip annunciation is provided on the main control panel.
The setpoints for low and high different ial pressure ala rms will be such that flows at
+/-10% of design flow wi ll be alarmed at the main control panel in the form of low or high differential pressure.
The differential pressures acr oss all HEPA filters are indicated on local control panels. High differential pressure across all HEPA filters is annunciated on the
main control panel and on local cont rol panels. Upset conditions will ther efore be identif ied to the control room operator allowi ng him to take a ppropriate action.
The auxiliary buildi ng exhaust flow rate and fuel handling building exhaust flow rate is continuously sensed by the exhaus t stack airflow me asuring equipment and recorded in a local control panel and by the plant computer.
2.j Filter trains are not designed to be removable from the building as an intac t unit. The size of the train precludes shipment o ff-site and there are no facilities for onsite
BYRON-UFSAR A1.52-2 REVISION 1 - DECEMBER 1989 disposal of the intact unit. The filter elements are removable and can be dis posed of through the solid radwaste system.
2.l Filter Housings
All of the auxiliary building and fu el handling building exhaust system filter housings are desig ned in accordance with ANSI N509-76. The hous ings are at negative pressure with respect to thei r surroundings and are located in auxiliary building general a rea which is a low airborne radiation environment. Any in-leakage from the general area will not ad versely affect Appendix I releases.
Hence, the housings we re not leak tested to the ANSI N509 requirements. Howev er, filter mounting leak tests were performed in accordance with ANSI N510-80.
The control room emergency m akeup air system filter housings are designed in acc ordance with ANSI N509-76.
The filter housings are at negative pressure with respect to their surroundings, and are located w ithin the control room boundary which is a habitable env ironment the same as the control r oom. Any in-leakage will be from the control room environment and, therefore, will not adversely affect the quality of that environment; hence, the housings were not leak tested to ANSI N509 requirements. Howev er, filter mounting frame tests were performed in accorda nce with ANSI N5 10-80. Some field welds may have been painted prior to performing the mounting frame l eakage test.
Ductwork All auxiliary building and fuel handling building exhaust system ductwork upst ream of the filter units is under negative pressure with respect to its su rroundings and is located in the same ar eas of the buildings served by the exhaust systems. An y in-leakage will be filtered prior to discharge to the atmosphere, hence, this ductwork was not be tested to ANSI N509 requirements.
However, prior to the final system turnover, the nonaccessible exhaus t system of the auxiliary building HVAC system and the fuel handling building exhaust system was operated and the ductwor k was visual ly and audibly checked for leaks. Le aks were sealed.
All control room emergency mak eup air system ductwork is located within the c ontrol room boun dary which is a habitable environment. Any du ctwork leakage will not adversely affect the habitab ility of the environment, hence, this ductwork was not tested to ANSI N509 requirements. Howev er, prior to final system turnover, the control room emergency makeup air system was operated and visual and audib le leaks in duct work were sealed.
BYRON-UFSAR A1.52-3 REVISION 15 - DECEMBER 2015 The design airflow q uantities for each s ystem were verified during testing, adjust ing and balancing of the systems.
Deviations of more than
+/- 10% of the design flow quantities were evaluated and any disposition was documented. For the auxiliary building exhaust system, minimum airflow quantities are limit ed to system requi rements while maximum airflow quantities a re limited to 110%
of the installed filter capacity. A ca librated orifice was used in lieu of
a gas flow totalizer f or determining leakage.
2.m (Deleted) (Note 1) 3.b (Deleted) (Note 1)
3.d Replacement HEPA filters sha ll meet the requ irements of ANSI N509-1989 in li eu of N509-1976.
3.e (Deleted) (Note 1) 3.h (Deleted) (Note 1)
3.i Replacement activate d carbon shall m eet the requirements of Table 5.1 of ANSI N509-1980 in lieu of ANSI N509-1976, which was replaced by ANSI N509-1980.
3.n Ductwork is designed, co nstructed, and tested in accordance with intent of Section 5.10 of ANSI N509-1976.
The longitudinal seams, howeve r, are either seal welded or mechanical lock type (Pittsburgh lo ck with sealant).
Silicone sealant is used as a permanent sealant for HVAC ductwork.
Fan peak pressure tests were not performed. For systems that have isolation devices, the fans are provided with high differential pressure t rips or high/low flow trips.
3.p Bubble tight iso lation and shutoff d ampers are provided only for the control room intakes. Two parallel blade isolation dampers in series are provided in the VA system nonaccessible system cha rcoal filter bypass.
4.b The space provided between components is 3 feet from the front (or rear) of t he components to the nearest obstacle (filter frame or other filter component).
This allows 3 feet of access b etween components.
4.c (Deleted) (Note 1)
4.d The periodicity of Emergency Makeup Unit testing is set in accordance with the Surveillance Frequency Control Program. 5.b Airflow distribution tests were perfor med to ensure that the airflow through any individu al filter element does not exceed 120% of the element's rated capacity.
BYRON-UFSAR A1.52-4 REVISION 15 - DECEMBER 2014 The VA system accessib le, nonaccessible and fuel handling building filters air capacity tests were performed to verify that maximum flow is not gr eater than 110% of filter rated capacity.
Airflow capacity tests w ere performed to ensure that the plenum flow rate does not e xceed 110% of filter rated capacity. The minimum flow rate is based on the exhaust flow rate needed to meet both ALARA and equipmen t qualification requirements.
Filtration unit airf low capacity tests w ere performed at the system design pressu re range correspondi ng to clean and dirty filter losses. Test s were performed at 1.25 tim es dirty filter conditions to verify system stability only.
Filter pressure losses for airflow c apacity tests were s imulated without filters in place.
5.c Silicone sealant was used as a permanent sealant for HVAC ductwork.
5.d The acceptance crite ria for bypass lea kage through the control room HVAC make-up ch arcoal adsorber is less than 1.0%.
- 6. Replacement activate d carbon shall m eet the requirements of Table 5.1 of ANSI N509-1980 in lieu of ANSI N509-1976, which was replaced by ANSI N509-1980.
6.b The acceptance c riteria for control room HVAC make-up charcoal adsorber lab analysis methyl iodide penetration is less than 2.0%.
The control room HVAC recirculation charcoal is tested to 48 fpm. Further discussions on t his subject can be f ound in Section 6.5 and Subsections 9.4.1.2 and 12.3.1.7.
Note 1: Exception to this sect ion is no longer required because the Regulatory Guide has bee n revised to eliminate the criteria to which exception was originally taken.
BRAIDWOOD-UFSAR A1.52-5 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.52 DESIGN, TESTING AND MAIN TENANCE CRITERIA FOR ENGINEERED-SAFETY-FEATURE ATMOSPHERE CLEAN UP SYSTEM AIR FILTRA TION AND ADSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR POWER PLANTS
The Licensee complies with Revision 2 of the regulatory position with the following c omments and exceptions keyed to paragraph numbers in Section C of the position:
1.e (Deleted) (Note 1) 2.a Entrained water droplets are not considered credible due to significant q uantities of ductwork with elbows. Water droplets, if present, will i mpinge on ducts and drop out of vertical duct risers as t he air enters exhaust plenums. However, t he auxiliary building exhaust system does contain pre filters which can se rve as demisters.
2.d (Deleted) (Note 1)
2.f The auxiliary buildi ng nonaccessible area exhaust filter system consists of t hree built-up filter trains (one standby) with a rated capacity of 63,000 cfm each. The system design flow is 62,730 cfm.
For maintenance purposes each train is divided into three banks with each bank sized for seven filters wide an d three high. Each train has a total of 63 HEPA filter elements.
The auxiliary building accessible ar ea exhaust filter system has a rated capacity of 189,000 cfm. The system design exhaust flow is 125,490 cfm. This system consists of four built-up fil ter trains (one st andby), and each train is divided into three ba nks. Each bank is sized for seven filters wide and three filters high. Each train has a total of 63 HEPA filter elements.
2.g ALL ESF filter syste ms have local cont rol panel airflow indication. In additi on, the flow rate of each of the stacks is recorded at the local control panel. The airflow rate through the control room em ergency makeup air filter units and the aux iliary building and fuel handling building exhaust ch arcoal booster fans is continuously sensed.
BRAIDWOOD-UFSAR A1.52-5a REVISION 8 - DECEMBER 2000 The differential pressure ac ross all of the ESF filter unit fans is indicated on lo cal control panels.
High and low differential press ure, and fan trip annunciation is provided on the main control panel.
The setpoints for low and high different ial pressure ala rms will be such that flows at +10% of design f low will be alarmed at the main control panel in the form of low or high differential pressure.
The differential pressures acr oss all HEPA filters are indicated on local control panels. High differential pressure across all HEPA filters is annunciated on the main control panel and on local cont rol panels. Upset conditions will ther efore be identif ied to the control room operator allowi ng him to take a ppropriate action.
The auxiliary buildi ng exhaust flow rate and fuel handling building exhaust flow rate is continuously sensed by the exhaus t stack airflow me asuring equipment and recorded in a local control panel and by the plant computer.
2.j Filter trains are not designed to be removable from the building as an intac t unit. The size of the train precludes shipment o ff-site and there are no facilities for onsite disposal of the intact un it. The filter elements are removable and can be dispos ed of through the solid radwaste system.
BRAIDWOOD-UFSAR A1.52-6 2.l Filter Housings All of the auxiliary building and fu el handling building exhaust system filter housings are desig ned in accordance with ANSI N509-76. The hous ings are at negative pressure with respect to thei r surroundings and are located in auxiliary building general a rea which is a low airborne radiation environment. Any in-leakage from the general area will not ad versely affect Appendix I releases.
Hence, the housings we re not leak tested to the ANSI N509 requirements. Howev er, filter mounting leak tests were performed in accordance with ANSI N510-80.
The control room emergency m akeup air system filter housings are designed in acc ordance with ANSI N509-76.
The filter housings are at negative pressure with respect to their surroundings, and are located w ithin the control room boundary which is a habitable env ironment the same as the control r oom. Any in-leakage will be from the control room environment and, therefore, will not adversely affect the quality of that environment; hence, the housings were not leak tested to ANSI N509 requirements. Howev er, filter mounting frame tests were performed in accordance with ANSI N510-80.
Some field welds may have been painted prior to performing the f ilter mounting frame leak tests.
Ductwork All auxiliary building and fuel handling building exhaust system ductwork upst ream of the filter units is under negative pressure with respect to its su rroundings and is located in the same ar eas of the buildings served by the
exhaust systems. An y in-leakage will be filtered prior to discharge to the atmosphere, hence, this ductwork was not tested to ANSI N509 requirements.
However, prior to the final system turnover, the nonaccessible exhaus t system of the auxiliary building HVAC system and the fuel handling building exhaust system will be operated and the ductwork was visually and audibly checked for leaks.
Leaks were sealed.
All control room emergency mak eup air system ductwork is located within the c ontrol room boun dary which is a habitable environment. Any du ctwork leakage will not adversely affect the habitab ility of the environment, hence, this ductwork was not tested to ANSI N509 requirements. Howev er, prior to final system turnover, the control room emergency makeup air system was operated and visual and audible leaks in the ductwork were sealed.
BRAIDWOOD-UFSAR A1.52-7 REVISION 7 - DECEMBER 1998 The design airflow quant ities for each system was verified during testing, adjus ting and balan cing of the systems. Deviations of more than
+/- 10% of the design flow quantities were evaluated and a ny disposition was documented. For the auxilia ry building exha ust system, minimum airflow quantities are limited to system requirements while maximum airflow qua ntities are limited to 110% of the installed filter capaci ty. A calibrated orifice was used in lieu of a gas flow totalizer for determining leakage.
2.m (Deleted) (Note 1) 3.b Initial duct heater perf ormance testing per ANSI N510-80 was not performed for the Cont rol Room Make-up Air Filter Unit heating coils. I nstalled heater capacity was verified using contr ol room HVAC system pre-operational test data. The revi ew of the test data demonstrated that the heaters can perform at their design capacity.
3.d Replacement HEPA filters sha ll meet the requ irements of ANSI N509-1989 in li eu of N509-1976.
3.e (Deleted) (Note 1) 3.h (Deleted) (Note 1) 3.i Replacement activate d carbon shall m eet the requirements of Table 5.1 of ANSI N509-1980 in lieu of ANSI N509-1976, which was replaced by ANSI N509-1980.
3.n Ductwork is designed, co nstructed, and tested in accordance with the intent of Section 5.10 of ANSI N509-1976. The long itudinal seams, ho wever, are either seal welded or mechani cal lock type (Pit tsburgh lock with sealant). Silicone seal ant is used as a permanent sealant for HVAC ductwork.
Fan peak pressure tests were not performed. For systems that have isolation devices, the fans are provided with high differential pressure t rips or high/low flow trips.
3.p Bubble tight iso lation and shutoff d ampers are provided only for the control room intake. Two parallel blade isolation dampers in series are provided in the VA system nonaccessible system cha rcoal filter bypass.
4.b The space provided between components is 3 feet from the front (or rear) of t he components to the nearest obstacle (filter frame or other filter component).
This allows 3 feet of access b etween components.
4.c (Deleted) (Note 1)
BRAIDWOOD-UFSAR A1.52-7a REVISION 15 - DECEMBER 2014 4.d The periodicity of Emergency Makeup Unit testing is set in accordance with the Surveillance Frequency Control Program. 5.b Airflow distribution tests were perfor med to ensure that the airflow through any individu al filter element does not exceed 120% of the element's rated capacity.
The VA system accessib le, nonaccessible and fuel handling building filters air capacit y tests were performed to
verify that maximum flow is not greater than 110% of the
BRAIDWOOD-UFSAR A1.52-8 REVISION 13 - DECEMBER 2010 filter rated capacit
- y. The minimum flow rate is based on the exhaust flow rate needed to meet both ALARA and equipment qualificat ion requirements.
Filtration unit airflow capacity tests were performed at the system design pressure r ange corresponding to clean and dirty filter issues. Th e midpoint filte r drop test was not performed. Tests were performed at 1.25 times dirty filter conditions to verify system stability only.
Filter pressures losses for airflow capacity tests were simulated without filters in place.
5.c Silicone sealant or other te mporary patching material was not used in the ESF filter hou sings. Silico ne sealant is used, however, as a permanent sealant for HVAC ductwork.
A sampling rate of less than 1 c fm was employed for testing filter systems l arger than 1000 cfm.
The acceptance criteria for the VA Nonaccessible (NAC) and Fuel Handling Building (FHB) Systems is per the VFTP.
The amount of leakag e bypassing the NAC HEPA filters on the standby train when deter mining NAC T otal System Bypass leakage will be t he amount measur ed when the train is on-line.
5.d The acceptance crite ria for bypass lea kage through the control room HVAC make-up ch arcoal adsorber is less than 1.0%. The acceptance criteria for the NAC and FHB Systems is per the VFTP.
The amount of leakage bypassing the NAC charcoal adsorbers on the standby tra in when determining NAC Total System Bypass le akage will be the amount measured when the train is on-line.
6.a(2) All carbon fur nished prior to 1985 as part of the original specification for atmospheric clean-up filtration units was tested to the req uirements of Table 5-1 of ANSI N509-1976. All replacement carbon or original carbon furnished in 1985 or later will be tested to the requirements of Table 5-1 of AN SI N509-1980 with the exception that the laboratory test f or methyl iodine penetration at 30
° C, 95% relative humidity is less than 1%.
BRAIDWOOD-UFSAR A1.52-8a REVISION 13 - DECEMBER 2010 6.a(3) Laboratory tests will be performed per t he Ventilation Filter Testing Program.
6.b The acceptance c riteria for control room HVAC make-up charcoal adsorber lab analysis methyl iodide penetration is less than 2.0%.
The control room HVAC recirculat ion charcoal is tested to 48 fpm. The acceptance criteria for the VA NAC and FHB System charcoal penetration of the standby train (service condition when VA System is in the emergency mode) is encompassed by the t esting performed at the rated flow with the train on-line.
The acceptance criteria for the NAC and FHB Systems is per the VFTP.
Further discussions on t his subject can be f ound in Section 6.5 and Subsections 9.4.1.2 and 12.3.1.7.
Note 1: Exception to this sect ion is no longer required because the Regulatory Guide has bee n revised to eliminate the criteria to which exception was originally taken.
B/B-UFSAR A1.53-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.53 APPLICATION OF THE SINGL E-FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION SYSTEMS
The Licensee complies with Revision 0 of the guidelines for application of the single failure criteria to nuclear power plant protection systems as discussed in Sub sections 7.1.2.11 and 8.1.10.
B/B-UFSAR A1.54-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.54 QUALITY ASSURANCE REQUIREMENTS FOR PRO TECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS The Licensee complies with Revision 0 of the regulatory guide position with the exception of s ome undocumented or unqualified coatings. (See Subs ection 6.1.2 for further information.)
B/B-UFSAR A1.55-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.55 CONCRETE PLACEMENT IN CATEGORY I STRUCTURES
The plant design conforms to t he regulatory position in Revision 0 with the follo wing exceptions:
- 1. ACI 301-72 specifies that the frequency for cylinder testing shall be two cylinde rs per 100 yards of concrete, tested at 28 days wi th a minimum of one set per day for each class of concrete.
The Licensee's posit ion is to use six cylinders per 150 yards of concrete, t ested at 7, 28, and 91 days, with a minimum of one set per day for each class of concrete. This exceeds the requirements of both ACI 318-77 and A CI 349-76.
The compliance with the requirements of this regulatory guide is discusse d in detail in Section B.1 of Appendix B.
- 2. ACI 301-72, Subsection 8
.5.3, requires t hat grouting be applied on the vertic al surfaces of construction joints. This requir ement has been r emoved from ACI 381-80. The requirements remain in effect even though the regulatory guide was withdrawn on J uly 8, 1981. The re gulatory position is now covered by one or mo re national standards.
B/B-UFSAR A1.56-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.56 MAINTENANCE OF WATER PURITY IN BOILING WATER REACTORS This regulatory guide is per tinent to BWRs only.
B/B-UFSAR A1.57-1 REVISION 16 - DECEMBER 2016 REGULATORY GUIDE 1.57 DESIGN LIMITS AND LOAD ING COMBINATIONS FOR METAL PRIMARY REACTOR CO NTAINMENT SYSTEM COMPONENTS The licensee complies wi th the regulatory po sition in Revision 0 of this guide wi th the following clarifications.
Piping penetration assemblies are design ed by the following guidelines:
- a. The portion of the prima ry containment penetration assembly which is part of the containment boundary, i.e., the penetration sleeve in its entire length (including the sleeve projecti on that forms an extension to the wall), is designed in a ccordance with Subsection NE, Secti on III of the AS ME Code, augmented by the applicable pr ovisions of Regula tory Guide 1.57.
- b. The portion of the prima ry containment penetration assembly which consists of the head fitting (flued head) and part of the proces s pipeline, is designed in accordance with Subs ection NB of the Code so as to satisfy stress requi rements for design conditions (NB-3112, NB-3221), normal and upset conditions (NB-3113.1, NB-3112.2, NB-32 22, NB-3223), emergency conditions (NB-3 224), faulted condit ions (NB-3113.4, F-1324.1, F-1324.6, Table F-1322), and testing conditions (NB-3226, NB-6222, NB-6322).
Part b of the Licensee's pos ition, which refers to the NB classification of the flued head and process pipe, is supported by NA-2134 of the Code and note 3 of Regulatory Guide 1.57.
Part b discussion is applicable to T ype 1 penetration type as described in UFSAR sections 3.8.2.1.3.1.1 and 3.8.2.1.3.2. All ot her head fittings are MC component and designed to appl icable NE requirements of ASME Section III as discussed in Pa rt a and UFSAR Section 3.8.2.1.3.2.
B/B-UFSAR A1.58-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.58 QUALIFICATION OF NUCLEAR POW ER PLANT INSPECTION, EXAMINATION, AND TESTING PERSONNEL The requirements of Regu latory Guide 1.58 ha ve been incorporated in Regulatory Guide 1.28, Revision 3. Regul atory Guide 1.58 was withdrawn on June 17, 1991.
The Licensee complies with the intent of Reg ulatory Guide 1.28 Revision 3, but applies it to ASME/ANSI NQA-1-1994.
B/B-UFSAR A1.59-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.59 DESIGN BASIS FLOODS FOR NUCLEAR POWER PLANTS
The plant design conforms to t he regulatory positions in Revision 2 as described in Section 2.4.
B/B-UFSAR A1.60-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.60 DESIGN RESPONSE SPEC TRA FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS The plant design conforms to t he regulatory positions in Revision 1 as described in Subsection 2.5.2.
B/B-UFSAR A1.61-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.61 DAMPING VALUES FOR SEISM IC DESIGN OF NUCLEAR POWER PLANTS
The plant design conforms to t he regulatory positions in Revision 0 as described in Su bsections 3.7.1.3 and 3.7.3.14, and 3.7.3.15 with the single exception of t he large piping sy stems (diameter greater than 12 inches) SSE conditio n value of 3% critical. A conservative value of 4% critical for the West inghouse reactor coolant loop configura tion has been justifie d by testing and has been approved by the N RC staff. The test results are given in WCAP-7921-AR, "Damping V alues of Nuclear Power P lant Components." The use of higher damping valu es, when justifi ed by documented test data, have been provided for in R egulatory Position C2.
B/B-UFSAR A1.62-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.62 MANUAL INITIATION OF PROTECTIVE ACTIONS
The Licensee complies with Revision 0 of thi s regulatory guide.
Refer to Section 7.3 and Subsections 8.1.11, 7.1.2.1.2, 7.1.2.12.1, and 7.2.1.1.2 for further information.
B/B-UFSAR A1.63-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.63 ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR LIGHT-WATER-COOLED NUC LEAR POWER PLANTS The plant design specification r equires the pene tration vendors to meet the requirements of Regu latory Guide 1.6 3, Revision 0, which was in eff ect at the time the construction permit application was docketed.
Regulatory Guide 1.
63 supplements IEEE 317-1972, the IEEE S tandard for "Electric Penetration Assemblies in Containment S tructures for Nuclear Po wer Generating Stations," and contains no specific testing recommendations. Regulatory Position C-4 does however add the quality assurance requirements of ANSI N45.2-1971 a nd ANSI N45.2.4-1972 to Section 8 (Required Data and Quality Control and Qua lity Assurance P rocedures) of IEEE 317-1972.
The design, construction, inst allation, and testing of the electrical penetration a ssemblies will be in accordance with the quality assurance requir ements of Regulatory Position C-4. See Subsections 3.11.2, 7.1.2.13 , and 8.1.12 for further information.
B/B-UFSAR A1.64-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.64 QUALITY ASSURANCE RE QUIREMENTS FOR THE DESIGN OF NUCLEAR POWER PLANTS The requirements in Regu latory Guide 1.64 ha ve been incorporated in Regulatory Guide 1.28, Revision 3. Regul atory Guide 1.64 was withdrawn on June 17, 1991.
The Licensee complies with the intent of Reg ulatory Guide 1.28 Revision 3, but applies it to ANSI/ASME NQA-1-1994.
B/B-UFSAR A1.65-1 REVISION 11 - DECEMBER 2006 REGULATORY GUIDE 1.65 MATERIALS AND INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS The Licensee complies with Revision 0 of Reg ulatory Guide 1.65, except for material and tensile streng th guidelines and supplemental inservi ce inspection (I SI) examinations.
Westinghouse has specifi ed both 45 ft-lb and 25 mils lateral expansion for control of fract ure toughness determined by Charpy-V testing, required by the ASME Boiler and Pr essure Vessel Code, Section III, Summer 1973 Addenda and 10 CFR 50, Appendix G (July 1973, Paragraph IV.A.4). These toughn ess requirements assure optimization of t he stud bolt materia l tempering operation with the accompanying reduction of the t ensile strength level when compared with p revious ASME Boiler and Pressure Vessel Code requirements.
The specification of both impact and maximum ten sile strength as stated in the guide results in unneces sary hardship in procurement of material without any addition al improvement in quality. The closure stud bol ting material is procured to a minimum yield strength of 130,000 psi and a minimum tensile strength of 145,000 psi. This s trength level is compatible with the fracture toughness requirements of 10 CF R 50, Appendix G (July 1973, Paragraph 1.C), although hig her strength level bolting materials are permitted by t he code. St ress corrosion has not been observed in reactor vessel clos ure stud bolting manufactured from material of th is strength leve
- l. Accelerated stress corrosion test data do exist for materials of 170,000 psi minimum yield strength exposed to marine water environments stressed to 75% of the yield strength (given in Reference 2 of the guide). These data are not consid ered applicable to Westinghouse reactor vessel closure stud bol ting because of the specified yield strength diffe rences and a less severe environment; this has been dem onstrated by yea rs of satisfactory service experience.
The ASME Boiler and Pressure V essel Code requirement for toughness for reactor vessel bol ting has preclud ed the guide's additional recommendation for te nsile strength limit ation, since to obtain the requir ed toughness levels, the tensile strength levels are reduced.
Prior to 1972, the Code required to 35 ft-lb toughness level whic h provided maximum t ensile strength levels ranging from approxima tely 155 to 178 ksi (Westinghouse
B/B-UFSAR A1.65-2 REVISION 16 - DECEMBER 2016 review of limited data - 25 heat s). After pub lication of the Summer 1973 Addenda to the Code and 10 C FR 50, Appendix G, wherein the toughness re quirements were modifi ed to 45 ft-lb with 24 mils lateral expansio n, all bolt material data reviewed on Westinghouse plants show ed tensile strengths of less than 170 ksi. Additional protectio n against the poss ibility of incurring corrosion effects is assured by:
- 1. Decrease in level of tensile strength comparable with the requirement of fracture toughness as described above.
- 2. Design of the re actor vessel studs, nuts, and washers, allowing them to be completely r emoved during ea ch refueling permitting visual and/or non destructive inspection in parallel with refueling operations to as sess protection against corrosion, as part of the inservice inspection program.
- 3. Design of the re actor vessel studs, nuts, and washers, providing protection against cor rosion by allowi ng them to be completely removed during each refueling and placed in storage racks on the c ontainment operating deck, as required by Westinghouse refuel ing procedures. T he stud holes in the reactor flange are sealed with special p lugs before removing the reactor closure.
Thus, the bolting materials and stud holes are never expo sed to the borated refueling cavity water.
- 4. Use of a manganese phosp hate surface treatment.
Use of Code Case 1605 does not constitute an issue between the NRC and Westinghouse i nasmuch as (a) no questions have been raised on this point in vendor's approved standard reference document discussions of this guide and (b) u se of this code case has been approved by the NRC v ia the guideline of Regulatory Guide 1.85.
Inservice inspection examinations of reactor pressure vessel bolting (closure head st uds, nuts, washers, etc.) are performed in accordance with t he methods specified in the station Ten Year Inservice Inspection (ISI) Plan which is manda ted through 10 CFR 50.55a. Volumetric ex amination of bolti ng is performed in accordance with the pr ocedures and qualifica tions of approved versions of ASME Section XI, D ivision 1, Appendix VIII, "Performance Demonstra tion for Ultrasonic Examination Systems" with Supplement 8 "B olts and Studs" mand ated through 10 CFR 50.55a. Further discussion of reactor coolant pressure boundary materials, inspection, a nd testing is in Sub sections 5.2.3 and 5.2.4.
B/B-UFSAR A1.66-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.66 NONDESTRUCTIVE EXAMI NATION OF TUBULAR PRODUCTS
The regulatory positions of the guide have been incorporated in Section III of the ASME code for tubular pro ducts intended for use in safety-relate d systems. This code is used to perform nondestructive examinati ons. Regulatory Guide 1.66 was withdrawn on September 28, 1977 becaus e it was no longer needed.
B/B-UFSAR A1.67-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.67 INSTALLATION OF OVERPRES SURE PROTECT ION DEVICES
The recommendations of this guide ar e included in th e ASME Boiler and Pressure Vessel Code, which is inc orporated by r eference in 10 CFR 50.55a. The re gulatory guide was wit hdrawn on April 15, 1983 because it was no longer needed.
B/B-UFSAR A1.68-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.68 INITIAL TEST PROGRAMS FOR WATER-COOLED REACTOR POWER PLANTS The Licensee complies with Revision 2 of thi s regulatory guide, as described in Chap ter 14.0, with the f ollowing exceptions:
Appendix A.2.b states "To the ex tent practical, testing should demonstrate control rod scram times at both hot zero power and cold temperature conditions, with flow and no-fl ow conditions in the reactor coolant syst em as required to bo und conditions under which scram might be required." A full spectrum of r od drop measurements was made for Byron Unit 1 at cold no-flow, hot no-flow, cold full-flow, and hot full-flow conditions.
Byron Unit 2 and Brai dwood Units 1 and 2 are identical to Byron U nit 1 with respect to the rod control system. Because of this, no a dditional design i nformation would be obtained by repea ting the entire sp ectrum of rod drop measurements that was original ly done for Byron Unit 1.
Consequently, the Licensee int ends to perform only the hot full-flow measurements for the remaining three units.
Appendix A.4.c states "Following initial cri ticality, licensee should conduct pseudo-ro d-ejection test to v erify calculational models and accident an alysis assumptions." The results of the Byron Uni t 1 pseudo-rod-ejection test confirmed the design predictions made for the ev ent within the accuracy of the testing procedure. Verifica tion of core design parameters for the remaining three units can be achieved through control rod worth measur ements and flux mapping at z ero power and during the power ascensi on phase. Conse quently, the Licensee does not intend to per form the pseudo-rod-ejection test for Byron Unit 2 and Braidwood Units 1 and 2 because no additional information will be provided with regard to core performance because of the d esign similarity.
Appendix A.4.t states "P erformance of natural circulation tests of the reactor c oolant system to confirm that the design heat removal capability exists or to verify that flow (without pumps) or temperature data are compatible to protot ype designs for which equivalent tests have been s uccessfully completed (PWR).
"As described in the B yron SER Secti on 5.4.3 the Licensee has referenced the natur al circulation testing w hich was performed at Diablo Canyon. The NRC staff and
B/B-UFSAR A1.68-2 REVISION 15 - DECEMBER 2014 Brookhaven National La boratory have reviewed the Diablo Canyon test results and found them acceptable. A p reliminary assessment of differences b etween Byron and Diablo Canyon that may affect boron mixing under n atural circulation h as been provided and indicates that the D iablo Canyon test re sults and supporting analysis satisfy the necessa ry requirements for Byron.
Byron/Braidwood Statio ns and Diablo Cany on Unit 1 have subsequently been compared in de tail to ascertain any differences between the plants t hat could potentially affect natural circulation flow and a ttendant boron mixing.
Because of the similarity between t he plants, the Licen see concluded that the natural circulation capa bilities would be simi lar, and therefore, the results of prototypical natu ral circulation cooldown tests conducted at Diablo Canyon w ould be representative of the capability at Byron/Br aidwood. The plant co mparison is further discussed in subsect ion 5.4.7. Based on the review of the similarities between Byron/Braidwood and Diablo Canyon, the NRC has concluded that Byron and Bra idwood have demonstr ated that the Diablo Canyon na tural circulation tests are applicable to Byron/Braidwood and that the y comply with the requirements of BTP RSB 5-1 (Reference 1
). Additionally, si mulator training for Byron reactor operators includes natural circu lation procedures training.
Appendix A.5.a state s "Determine that power reactivity coefficients (PWR) or power vs. flow character istics (BWR) are in accordance with design values (25%, 50%, 75%, 100%)."
Per recommendations of Westingho use, Byron N SSS vendor, the Licensee intends to perform this tes ting at the 30%, 50%, 75%, and 90% power ascension testing plateaus. T hese testing plateaus correspond to those pr eviously listed in Tab le 14.2-82 for the Power Reactivity Coefficient Measurement Startup Test.
Appendix A.5.e states "P seudo-rod-ejection test to validate the rod ejection acc ident analysis."
The results of the Byron Uni t 1 pseudo-rod-ejection test confirmed the design predictions made for the ev ent within the accuracy of the testing procedure. Verifica tion of core design parameters for the remaining three units can be achieved through control rod worth measur ements and flux mapping at z ero power and during the power ascensi on phase. Conse quently, the Licensee does not intend to per form the pseudo-rod-ejection test for Byron Unit 2 and Braidwood Units 1 and 2 because no additional information will be provided with regard to core performance because of the d esign similarity.
Appendix A.5.h states "C heck rod scram times from data recorded during scrams that o ccur during the startup test phase to determine that the s cram times remain within allowable limits." During power asc ension testing, the Licensee does not intend to formally instrument the rod po sition indicat ion system
B/B-UFSAR A1.68-3 REVISION 6 - DECEMBER 1996 for reactor trip review purpos es because this would require removal of the r od position indi cation from service.
This would be a violation of Technical Spec ifications. For this reason the Licensee believes that this requ irement applies to BWRs only.
Appendix A.5.i states "D emonstrate capability and/or sensitivity, as appropriate for t he facility design of incore and excore neutron flux instrum entation, to detect a control rod misalignment equal to or less than the technic al specification limits (50%, 100%) (PWR)."
An evaluation of instrum entation response to a misaligned control rod will be performed during the Byron Unit 1 fl ux asymmetry startup test conducted at 50% power. The Licensee does not intend to perform this test on Byron Unit 2 or Braidwood Units 1 and 2 because no additio nal design confirmation will be obtained due to identical core co nfigurations and system designs. Also, in accordance with a recommendat ion from the NSSS vendor, the Licensee does not intend to perf orm this testing at 100% power. Although the technical specifications provide relief from the requirements of certain technical specificat ions when performing physics tests below 85%
power, creating a control rod misalignment at 100%
power does not fall wit hin this special test exclusion. As a resul t, technical specifica tion limits regarding rod insertion and/or peaking factors may be exceeded.
Appendix A.5.j states "Verify that plant performance is as expected for rod runback and partial scram."
The Licensee asserts t hat these particular e vents are applicable to BWRs only, and theref ore will not be performed.
Appendix A.5.ff states, "Demonstrate or verify that important ventilation and air-cond itioning systems, including those for the primary containment and steamline tunnel, continue to maintain their service areas with in the design limits (50%, 100%)."
The Licensee asserts that the reference to steamline tunnel ventilation applies only to boiling wa ter reactors, in which case this system would be safety-related up to th e turbine stop valves. For B yron/Braidwood, this steam line tunnel ventilation system is non-safety-related and, th erefore, no such demonstration will be performed.
Appendix A.5.gg states " If appropriate for t he facility design, conduct tests to determi ne operability of eq uipment provided for anticipated transient without scram (ATWS), if not previously done (25%)."
The initial test program did not include this testin g because, at that time, the facility design did not include an ATWS mitigation system (AMS). Subsequen tly, the facility desi gn was changed to include an AMS, as discussed in subsection 7.7.1.21. AMS was tested during implem ention of the design change.
B/B-UFSAR A1.68-4 REVISION 6 - DECEMBER 1996 Appendix A.5.kk states " Demonstrate that the d ynamic response of the plant is in accordance w ith design for t he loss of or bypassing of the feedwater heater(s) from a credible single failure or operator er ror that would res ult in the most severe case of feedwater temper ature reduction (50%, 90%)." As described in Subsecti on 15.1.1.2, the trans ient resulting from the simultaneous isola tion and bypass of a high-pressure feedwater heater is the most s evere heater i solation/bypass event. This accident yields a reduction in feedwater temperature of 18°F and a new specific e nthalpy of 399.83 Btu/lb
- m. This event is less severe than the transient resulting from a 10% step load increase provided the steam gene rator inlet temperat ure does not decrease by more than 55
°F and the specific feedwater enthalpy is greater than 342.06 Btu/lb
- m. Since the heater isolation/bypass accident satisfies these criteria, the simultaneous isolation and bypass of a heater event is boun ded by a 10% step increase in load. The 10%
load increase is tested as desc ribed in Table 14.2-88. Therefore, a feedwater heater bypa ss test as described by Appendix A.5.kk w ill not be performed.
Appendix A.5.mm states " Demonstrate that the d ynamic response of the plant is in accord ance with design for the case of automatic closure of all main st eam line isolation valves. For PWRs, justification for conducting t he test at a lower power level, while still demonstrat ing proper plant r esponse to this transient, may be su bmitted for NRC staff review (100%)."
The Licensee does not intend to perform this test because the closure of all main st eam isolation valves will result in a turbine trip per Byron S ubsection 15.2.4.
The generator/turbine trip test will be pe rformed at 100% power and is a more severe transient. The Licens ee will further perform a turbine trip at about 25% power on Byron Unit 1 with the turbi ne bypass valves closed and disab led. This will further verify p lant dynamic response. The combina tion of these two trip tests will verify the transient response of the plant and the capability of the secondary side decay heat removal systems to c ope with these transients. The disabling of the turbine bypass system will restrict that capability to the steam generator PORVs and auxiliary feedwater systems (i.e
., the safety-related systems) and will demonstrate their dynam ic capability.
The performance of the MSIV test would be re dundant and woul d provide no additional information regarding plant response or capability.
In effect, performan ce of this test wo uld only result in unnecessary cycling of this equipment.
A test program has been established to ensure that all structures, systems, and compone nts will satisfa ctorily perform their safety-related f unctions. This test program provides additional assurance that the plant has been p roperly designed and constructed and is r eady to operate in a manner that will not endanger the health and safety of the public, th at the procedures
B/B-UFSAR A1.68-5 REVISION 15 - DECEMBER 2014 for operating the plant safely have been evalu ated and have been demonstrated, and that the plant and p rocedures are fully prepared to operate the faci lity in a safe manner.
The test program includes simulation of equipment failures and control system malfuncti ons that could reasonably be expected to occur during the plant l ifetime. The test p rogram also includes testing for interactions such as the performance of interlock circuits in the reactor protection system.
It also determines that proper permissive and prohibit function s are performed.
Care is taken to ensure that redundant chann els of the equipment are tested independently.
The initial startup test ing, conducted after the fuel loading and before commercial operat ion, will confirm th e design bases and demonstrate, where p ractical, that the p lant is capable of withstanding the anticipated tra nsient and postu lated accidents.
A detailed description of the test progr am is provided in Chapter 14.0.
References
- 1. NRC Letter, "Byron Station U nits 1 and 2 and Braidwood Station Units 1 and 2, Natural Circulation Cooldown," dated November 4, 1988.
B/B-UFSAR A1.68.1-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.68.1 PREOPERATIONAL AND INITI AL STARTUP TESTING OF FEEDWATER AND CONDENSATE SYSTEMS FOR BOILING WATER REACTO R POWER PLANTS
This guide is pertin ent to BWRs only.
B/B-UFSAR A1.68.2-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.68.2 INITIAL STARTUP TEST PRO GRAM TO DEMONSTRATE REMOTE SHUTDOWN CAPABILI TY FOR WATER-COOLED NUCLEAR POWER PLANTS
The Licensee complies with the position in Revision 1 of this regulatory guide. Refer to Table 14.2-86 for additional information.
B/B-UFSAR A1.68.3-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.68.3 PREOPERATIONAL TESTING OF INSTRUMENT AND CON TROL AIR SYSTEMS
This guide, which replaces R egulatory Guide 1.80, is not applicable to Byron and Braidwood. Refer to the discussion of Regulatory Guide 1.80.
B/B-UFSAR A1.69-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.69 CONCRETE RADIATION SHIELDS F OR NUCLEAR POWER PLANTS The Licensee complies wi th the regulatory po sition in Revision 0 of this guide. Concre te radiation shielding is discussed in Subsection 12.3.2.
B/B-UFSAR A1.70-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.70 STANDARD FORMAT AND CONT ENT OF SAFETY ANALYSIS REPORTS FOR NUCLEAR POWER PLANTS The FSAR is written in accordance wi th the content and format set forth by Regulatory Guide 1.70, Revision 2, which was the current revision. The content and format have b een maintained in the UFSAR and its updates.
However, as part of the ongoing effo rt to improve the quality of the UFSAR, the guideli nes provided in Nu clear Energy Institute (NEI) 98-03, "Guidelines for U pdating Final Sa fety Analysis Reports," Revision 1, June 199 9, as endorsed by NRC Regulatory Guide 1.181, "Content of the Updated Final Saf ety Analysis Report in Accordance with 1 0CFR50.71(e)," Revis ion 0, September 1999, are used to further improve the content of the UFSAR.
While the UFSAR will continue to follow the gene ral organizational recommendations, i.e., format, specified in this Regulatory Guide, the reorganization options described in NEI 98-03 will be used to simplify infor mation contained in the UFSAR to improve its focus, clarity, and maintainability.
B/B-UFSAR A1.71-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.71 WELDER QUALIFICATION FOR AREAS OF LIMI TED ACCESSIBILITY
The Licensee maintains that limited-accessibil ity qualification or requalification d escribed in Revision 0 of Regulatory Guide 1.71 exceeds ASME Section III and IX requirements and is an unduly restrictive and unnecessary requirement.
Acceptability of welds will be determined by requ ired examination
- s. Multiple production welds of similar comp onents in the shop will be subjected to close c ontrol and supervision achieving the same purpose as the g uide. See Sub sections 5.3.1.4 a nd 5.2.3 for further information.
B/B-UFSAR A1.72-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.72 SPRAY POND PLASTIC PIPING
This regulatory guide do es not apply to this application, since the Byron/Braidwood de sign does not util ize spray ponds.
B/B-UFSAR A1.73-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.73 QUALIFICATION TESTS OF ELECTRIC VALVE OP ERATORS INSTALLED INSIDE THE CONTAINMENT OF NUCLEAR POWER PLANTS This regulatory guide indicates the NRC accept ance (with certain qualifications) of t he requirements of I EEE 382-1972, "IEEE Trial-Use Guide for Type Test of Cla ss I Electric Valve Operators for Nuclear Power Gene rating Stations."
The Licensee complies wi th the objectives se t forth in Revision 0 of this regulato ry guide as indicated in Sub sections 6.2.4.2 and 8.1.13.
B/B-UFSAR A1.74-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.74 QUALITY ASSURANCE TE RMS AND DEFINITIONS
The requirements in Regu latory Guide 1.74 ha ve been incorporated in Regulatory Guide 1.28, Revision 3. Regul atory Guide 1.74 was withdrawn on Sep tember 1, 1989.
The Licensee complies with the intent of Reg ulatory Guide 1.28 Revision 3, but applies it to ANSI/ASME NQA-1-1994.
B/B-UFSAR A1.75-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.75 PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS
NSSS Scope The commitment to comply with the intent of the requ irements in Revision 2 of this guide is pr esented in Subsect ions 7.1.2.2.1 and 7.1.2.2.2.
Non-NSSS Scope
Physical independence of redundant electric systems is discussed in Subsections 8
.1.14 and 8.3.1.4, respectively.
The Licensee complies with the requirements of this guide with the exceptions and/or cl arifications to the regulatory positions identified and j ustified below:
Regulatory Position C1 Section 3, "Isolatio n Device" should be supplemented as follows: "(Interrupting devic es actuated only by fault current are not consider ed to be isolation devices within the context of this document.)" Licensee's Position
Interrupting devices actuated only by fault current may be used as isola tion devices p rovided that the requirements of IEEE 384
-1977, including the change (fuses are accep table isolation devi ces in dc power circuits) as proposed by the IEEE-PES-NPEC-SC6.5 Working Group in their September 7, 1988 letter, are met.
Justification of Lic ensee's Position
There is no technica l justification for precluding the use of Class 1E circ uit breakers or Class 1E fuses actuated only by fault or over load current as circuit
interrupting or isolation devices. (For further discussion of this subject, see S&L letter to the NRC dated December 21, 1 978 and NRC's resp onse dated March 28, 1979.)
B/B-UFSAR A1.75-2 REVISION 6 - DECEMBER 1996 Byron/Braidwood Design Although the Licensee believ es that a single circuit breaker or fuse (actuated by fault c urrent only) provides adequate isolation, the Byro n/Braidwood design will incorporate the follow ing additional features to further ensure isolation and thus satisfy NRC concerns.
The Licensee (where prac tical) will provide two interrupting devices (in serie s) actuated only by fault current. These two interrup ting devices will be:
- 1) Class lE qualified and
- 2) coordinated with their upstream interrupting de vice; breakers will be periodically tested to verify coordina tion. Periodic testing of fuses, in dc power circuits, to verify coordination is not required, provided that each fuse is tested (for example, resistance measur ement) to verify overcurrent protection as designed. In lieu of periodic testing, a documented Periodic Inspection and Maintenance procedure shall be implemented which will ensure:
- that the proper size and type of fuse is installed,
- that the fuse shows no physical sign of deterioration, and
- that the fuse connections are tight and clean.
Any remaining non-Cl ass lE loads (no t utilizing two interrupting devices) will be tripped from t he Class lE buses with a Safety Injectio n coincident with loss of offsite power signal.
The cables wh ich supply non-Class lE loads from redundant Class lE buses are routed through separate raceways.
Regulatory Position C2 Section 3, "Raceway".
Interlocked arm or enclosing cable should not be construed as a "raceway." Licensee's Position Although not a "raceway" in the same sense as a conduit or cable tray, recognition of and design credit for t he additional protection provided by t he metallic jacket of interlocked armored cable should be included in the regulatory guide.
Use of armored cable, in lieu of th e separation distances stated in the regulatory guide, should be perm itted when justi fied by specific testing and/or analysis, as providing the requ ired degree of protection for C lass lE circuits again st specific credible hazards.
B/B-UFSAR A1.75-2a REVISION 6 - DECEMBER 1996 Justification of Lic ensee's Position There is no technical ju stification for preclu ding the use of armored cable, in lieu of separation dis tances, to provide adequate isolation between Class 1E and non-Class 1E circuits and between redundant Class 1E circuits, when sh own to be adequate by specific testing and/or analysis.
Regulatory Position C6 Analyses performed in accordance with Sections 4.5(3
), 4.6.2, and 5.1.1.2 should be submit ted as part of the Safety Analysis Report and should identify those circuits installed in accordance with these sections.
B/B-UFSAR A1.75-3 REVISION 6 - DECEMBER 1996 Licensee's Position The referenced analysis, whe n performed to justi fy deviation from specific requirements of standard IEEE 384-1974, shall be prepared on a case-by-case basis, shall be doc umented and be on permanent file, available for NRC review, but wi ll not be an integral part of the saf ety analysis report.
Justification of Lic ensee's Position The Licensee's position is consistent with t hat taken for other plant design records; e.g., routine design c alculations, design document revisions, etc.
Regulatory Position C7
Non-Class lE instrumentation and control circuits should not be exempted from the provis ions of Sect ion 4.6.2.
Licensee's Position
Low energy non-Class lE instrumentation and control circuits are not required to be p hysically separated or electrically isolated from "associated" circuits pro vided (a) the non-Class lE circuits are not routed with "associated" circuits of a redundant division, and (b) they are analyzed to demon strate that Class lE circuits are not degraded below an acceptable level. As part of the analysis, considerat ion shall be given to the potential energy and identific ation of the circuits involved.
Justification of Lic ensee's Position
The Licensee's position is consistent with t he industry consensus position regarding required se paration between non-Class lE circuits and "associated" circui ts, taken in the 1977 and 1981 revisions to IEEE-38 4, Section 4.6.1(4).
Regulatory Position C8
Section 5.1.1.1 should n ot be construed to imply that adequate separation of redund ant circuits can be achieved within a confined space such as a cable t unnel that is effectively unventilated.
Licensee's Position
Adequate separation of redundant Class lE circuits can be achieved in areas of the plant t hat are effectively unventilated.
Justification of Lic ensee's Position There is no technical ju stification for precludi ng the routing of redundant Class lE cir cuits through areas of the plant that may be "effectively unventilated" provided that adequate physical
B/B-UFSAR A1.75-4 REVISION 6 - DECEMBER 1996 separation is provided between redundant circu its and appropriate thermal derating factors for such ci rcuits have been incorporated into the plant design.
Regulatory Position C9 Section 5.1.1.3 shou ld be supplemented as follows:
"(4) Cable splic es, in raceways, should be prohibited." Licensee's Position Cable splices, either wi thin raceways or at the interface of raceways and equipment, etc., are permitted provided they are intended by the plant design as indicated on the design documents.
Justification of Lic ensee's Position There is no technical ju stification for preclu ding the use of cable splices within r aceways or at their interfaces with equipment, etc., provided that t hey are an integ ral part of the plant design as indicated on the design documents.
Regulatory Position C10 Section 5.1.2, the phrase "at a sufficient n umber of points" should be understood to mean at intervals not to exceed 5 feet throughout the e ntire cable length. Als o, the preferred method of marking cable is color coding.
Licensee's Position
Cable installed in exposed Class lE and "associated" circuit raceways shall be identi fied in a manner of sufficient durability and at sufficient intervals to f acilitate initial verification that the installation is in co nformance with the separation criteria. Methods of providing the justificatio n, other than color coding of the cable jacket, are acceptable.
Justification of Lic ensee's Position
There is no technical ju stification for requir ing the intervals of identification of such ca bles to not exceed every 5 feet throughout the e ntire cable length. Cab le system designs employing less frequent identification intervals and that provide for verification that the installation is in conformance with the separation criteria are acceptable.
Use of cable jacket color coding alone, as a met hod of providing cable identification, may not be as effective as alternative methods. Other methods, e.g., using unique cable identification number with a segregation code, could be a m ore positive method of facilitating verifi cation that the ca ble installation is in accordance with the design separatio n requirements.
B/B-UFSAR A1.75-5 REVISION 6 - DECEMBER 1996 Regulatory Position C11 Section 5.1.2 should be supplemented as follows: "The method of identification used should be simple and should preclude the need to consult any r eference material to dis tinguish between Class 1E and non-Class 1E circu its, between non-Class 1E circuits 'associated' with different redundant Class 1E systems, and between redundant Cl ass 1E systems."
Licensee's Position The method of initial in stallation verification need not preclude consultation of refe rence documents.
Justification of Lic ensee's Position
There is no technical justificat ion for precluding use of reference documents du ring the installation verification process (e.g., use of design documents and insta llation records). A system based upon the use of such refere nce documents could be a most effective check that a cable is installed in accordance with the design documents and in a raceway of compati ble segregation assignment.
Regulatory Position C12
Pending issuance of other acceptable criteri a, those portions of Section 5.1.3 (exclusi ve of the Note fol lowing the second paragraph) that permit t he routing of cables through the cable spreading area(s) and, by implic ation, the contr ol room, should not be construed as acce ptable. Also, Secti on 5.1.3 should be supplemented as foll ows: "Where feasible, redundant cable spreading areas shou ld be utilized." Licensee's Position Power cables installed in dedicated solidly enclosed metallic raceways in air (e.g., rigid steel conduit or solid cable trays with solid flush covers), ma y be routed thro ugh those areas designated as "cable spreading areas," w here justified by analysis or other suitable means.
Justification of Lic ensee's Position There is no technical justificat ion to preclude the routing of power cables through cable spreading a reas when they are installed in such a manner to present no hazard to other cabling, generally of a lower energy level, w ithin the area.
Regulatory Position C14 Section 5.2.1 should be supplemented as follows: "And should have independent air supplies."
B/B-UFSAR A1.75-6 REVISION 6 - DECEMBER 1996 Licensee's Position Redundant standby generating u nits shall be placed in separate safety class str uctures and shall be p rovided with separate ventilation and combus tion air systems.
Justification of Lic ensee's Position
The Licensee's position is an interpretation of what is believed to be the intent of the Regulatory Position.
Regulatory Position C15
Where ventilation is required, the sep arate safety class structures required by Section 5.3.1 s hould be served by independent ventilation systems.
Licensee's Position
Redundant batteries shall be housed in separate safety class structures, i.e., sepa rate from one another, not necessarily separate from everything else within its own safety division.
For example, a battery may be placed in the same safety class structure as the switchg ear for that division.
Justification of Lic ensee's Position
The Licensee's position is an interpretation of what is believed to be the intent of the Regulatory Position.
Regulatory Position C17
Regulatory Guide Position on S ection 4.6.1 " Separation from Class lE Circuits," of IEEE Std 384 (1974)
By not modifying Secti on 4.6.1 of IEEE S td 384 (1974) in a regulatory position, the regulat ory guide has en dorsed it as stated in the IE EE standard.
Licensee's Position
There is no justification for pr ecluding the use of technically acceptable analysis to justify, on a c ase-by-case basis, exceptions to the generally stat ed criteria for separation of non-Class lE circuit s, from Class lE c ircuits. When such analysis demonstrates th at the following req uirements are met, the non-Class lE circuits involv ed need not be c lassified as "associated" circuits.
For specific cases, where cable terminat ion or routing arrangements (e.g., ca bles leaving cable trays in free air entering equipment or passing th rough conduit sl eeves in walls) limit the available se paration distances betwe en non-Class lE and Class lE cables, to less than the mi nimum separation applicable
B/B-UFSAR A1.75-7 REVISION 6 - DECEMBER 1996 to redundant cables in raceways, such lesser separations are permitted provided that a documented ana lysis is performed to demonstrate that:
- a. the non-Class lE circuits are not routed with Class lE circuits of a redundant division or circuits "associated" with a re dundant division, and
- b. the Class lE circuits in volved are not degraded below an acceptable level.
The analysis will include consid eration of the p otential energies of the circuits involved; the physical a nd electrical isolation (i.e., barriers) provided for the circuits by the cable insulation, the cable sh ielding, and the cable jacketing systems; the degree of environmental qual ification and fire retardant characteristics of the c ables; and the potential for hazards in the specific area involved.
Justification of Lic ensee's Position
The Licensee's position is consistent with t he industry consensus technical position stated in the 1977 revision to IEEE-384, Section 4.6.1(3).
B/B-UFSAR A1.76-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.76 DESIGN BASIS TOR NADO FOR NUCLEAR POWER PLANTS
The plant design conforms to t he regulatory position in Revision 0 as described in Section 3.3.
B/B-UFSAR A1.77-1 REVISION 12 - DECEMBER 2008 REGULATORY GUIDE 1.77 ASSUMPTIONS USED FOR E VALUATING A CONTROL ROD EJECTION ACCIDENT FOR PRESSURIZED WATER REACTORS (Refer to Subsections 15.4.7 and 15.4.8.3 for details of this analysis.)
Westinghouse methods and crite ria are documented in WCAP-7588 Revision 1A which has been rev iewed and accepted by the NRC.
The results of their ana lyses show compliance with the Regulatory Position given in Sections C1 and C3 of Regula tory Guide 1.77, Revision 0. However, they take exception to Regulatory Guide Position C2 which implies that t he rod ejection accident should be considered as an emergency co ndition. Westin ghouse considers this a faulted condi tion as stated in ANSI N 18.2, "Nuclear Safety Criteria for the Design of Stati onary Pressurized Water Reactor Plants." Faulted cond ition stress limits wi ll be applied for this accident.
This guide, although u sed in the original plant design, has been superseded by Regulatory Guide 1
.183, "Alternative Radiological Source Terms for Evalu ating Design Basis Acc idents at Nuclear Power Reactors".
B/B-UFSAR A1.78-1 REVISION 13 - DECEMBER 2010 REGULATORY GUIDE 1.78 ASSUMPTIONS FOR EVALUA TING THE HABITAB ILITY OF A NUCLEAR POWER PLANT CONTROL ROOM DURING A POSTULATED HAZARDOUS CHEMICAL RELEASE
The Licensee complies with Revision 0 of thi s regulatory guide.
Refer to Section 2.2 and Sub section 6.4.1 for further information.
As allowed by paragr aph C.4 of Regulatory Guide (RG) 1.78, Revision 0, "The toxicity limits sho uld be taken from appropriate authoritative sources."
NUREG/CR-6624 is considered an appropriate authoritative resource and, ther efore, the toxicity limits contained within may be used for periodic toxic gas surveys in place of those cont ained in RG 1.78, Revision 0.
B/B-UFSAR A1.79-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.79 PREOPERATIONAL TESTING OF EMERGENCY CORE COOLING SYSTEMS FOR PRESSURIZED WATER REACTORS The Licensee complies wi th the requirements in R evision 1 of this guide. The containment spray system is tested up to the containment isolation valve while taking suction from the refueling water storage tank. In testing the RHR system under the recirculation conditions, the containment sumps are filled with cold water up to elevation 376 feet 9 i nches. The adequacy of NPSH available under postaccident recircu lation conditions to the RHR pumps is correct ed for water elevati on, temperature, and run out flow through the con tainment spray pumps.
Procedures to verify operability of the ECCS pum ps establish proper flow-requirements during flow tests c onducted at cold conditions. The capability of t he pumps to deliver required flows under accident c onditions has been verif ied by analysis to preclude any unnecessary thermal shock damage at hot operating conditions. Flow capa bilities were verified using data obtained from unplanned a nd planned safety inje ction actuatio n performed during the testi ng program. Check val ve operability has been evaluated to guideli nes and criteria e stablished in Table 14.2-34. See Chapter 14.0 f or further discu ssion of preoperational testing.
B/B-UFSAR A1.80-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.80 PREOPERATIONAL TESTING OF INSTRUMENT AIR SYSTEMS
The air systems in the Byron/Braidwood desig n are designated Safety Category II, Quality Group D. As non-safety-related equipment, the a ir system does not come under the provision of Regulatory Guide 1.80, Revision 0. This reg ulatory guide was renumbered and r eissued as Regulatory Gu ide 1.68.3. Regulatory Guide 1.80 was withdrawn on April 20, 1982.
B/B-UFSAR A1.81-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.81 SHARED EMERGENCY AND SHUTDOWN ELECTRIC SYSTEMS FOR MULTIUNIT NUCLEAR POWER PLANTS The Byron/Braidwood desi gn complies with the requirements in Revision 1 of this regulatory guide (which indicates the acceptable methods of complian ce with General De sign Criterion 5). The independence of each un it's onsite el ectrical systems are further discussed in Subsection 8.1.15.
The Licensee believes that the intent of Reg ulatory Guide 1.81 (Position C.1) was to disallow "normal" sharing of d-c systems, not to disallow the te mporary connection of one d-c bus to a source in the other un it during periods of testing and/or maintenance. Provisions for administratively controlled manually
actuated, interconnectio ns between the nonredundant Class 1E d-c bus for each unit improves the overall r eliability and availability of the d-c systems by allowing a means for manually providing power to a d-c bus at a time when it would otherwise have to be out of se rvice (e.g., to perf orm a battery discharge cell, etc.) That th is was the intent is evident from the "Discussion," in Regulat ory Guide 1.81, Part B, second paragraph, first sentence, which reads as follows:
"Sharing of onsite power systems at multi-unit power plant sites generally results in a reduction in the number and capacity of the onsite power sources to leve ls below those r equired for the same number of units loc ated at separate sites." The "interconnection" pr ovided in the Byron/Br aidwood design does not result in a reduction in eit her the number or the capacity of the d-c power source
- s. . ., i.e., the number and capacity of the d-c power sources for each of the two units are exactly the same as they would be if the units were located at separate sites.
The interconnection between each Uni t's Class 1E 125-Vdc systems, via the cross-tie, is limited by procedu ral and administrative controls. These controls ensure that combinations of maintenance and test operations will not preclud e the systems capabilities to supply power to the ESF d-c loads. The crit eria specifying the allowable combinations of maintenance and test operations will be governed by the plant technical specificatio ns. Coordination between unit operations required during main tenance and testing will be governed by admi nistrative controls.
B/B - UFSAR A1.82-1 REVISION 12 - DECEMBER 2008 REGULATORY GUIDE 1.82 WATER SOURCES FOR LONG-TERM RECIRCULATION COOLING FOLLOWING A LOSS-OF-COOLANT ACCIDENT The suction screens to the containment recirculation sumps have been replaced as part of the activities to respond to NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors". The replacement filters have an opening size of 1/12 inch and have been designed to comply with NRC regulations that have been published in support of NRC Generic Letter 2004-02.
The Byron and Braidwood plant design conforms to the requirements of regulatory Guide 1.82 Revision 3. Compliance with the regulatory position of this guide is discussed below:
- 1. PRESSURIZED WATER REACTORS 1.1 Features Needed To Minimize the Potential for Loss of NPSH The ECC sumps, which are the source of water for such functions as ECC and containment heat removal following a LOCA, should contain an appropriate combination of the following features and capabilities to ensure the availability of the ECC sumps for long-term cooling. The adequacy of the combinations of the features and capabilities should be evaluated using the criteria and assumptions in Regulatory Position 1.3.
1.1.1 ECC Sumps, Debris Interceptors, and Debris Screens 1.1.1.1 The Byron and Braidwood Stations containment recirculation sumps include two separate sumps, fully redundant, each servicing one train of the Emergency Core Cooling System (ECCS). The replacement screens for each sump are sized to the full design basis debris load.
1.1.1.2 Two redundant sump pits are physically separated from each other and are protected from high energy piping by the solid steel cover of the trash rack structure.
The trash rack is a structure made of tube steel and angle steel supports, with side vertical stainless steel grating and a solid checkered plate cover. The top of the trash rack structure is approximately four (4) ft above the containment floor elevation B/B - UFSAR A1.82-2 REVISION 12 - DECEMBER 2008 of 377 ft. Design loads for the trash rack have been determined to address hydrostatic pressure, drag force, and debris impact.
Dynamic effects due to design basis high energy line breaks need not be considered in the structural qualification of the trash rack, and dynamic effects from pipe breaks considered in the vicinity of the trash rack structure need not be considered. An additional load due to lead shielding has also been considered to address lead shielding activities during refueling outages. (Reference 1)
The checkered plate located on top of the trash rack structure is not permanently attached to the supporting framing members.
The physical configuration of these plates provides adequate lateral restrain for the applied seismic self-weight excitation force and horizontal drag force. Furthermore, the vertical OBE and SSE accelerations are both lower than 1.0. Therefore, the dead weight of the checkered plate is adequate to resist uplift during a seismic event. Operation of the ECCS in the Recirculation Mode is assumed only in the design basis analysis for a LOCA.
1.1.1.3 The containment recirculation sumps are located below the containment floor elevation of 377 ft. This is the lowest floor elevation in containment, exclusive of the reactor vessel cavity. This maximizes the depth of the pool water above the sump screens.
A trash rack structure, vertical grating on the perimeter and checkered plate on top, protects the openings for both sumps.
The sump suction pipe is located inside the sump pits. Each sump pit has a concrete slab ceiling with three openings that allow water to enter the pit. The concrete slab provides further protection for the screens and sump suction pipe. A debris interceptor plate is installed at each grating sector to prevent larger debris from accessing the sump pits.
1.1.1.4 The screens for each recirculation sump have been tested under design basis loading conditions. The trash rack will prevent debris sliding along the floor from reaching the screens.
1.1.1.5 The containment recirculation sumps are located between the Primary Shield wall and the Secondary Shield wall; this area is normally referred to as Inside Missile Barrier (IMB). Due to this location, the water leaking directly from the RCS piping, or connected piping located in the same area, does not flow B/B - UFSAR A1.82-3 REVISION 12 - DECEMBER 2008 through any restricted path. The same argument applies for Containment Spray wash down water that falls into the IMB area through the Steam Generators enclosures, the Reactor Coolant Pumps enclosures and the Pressurizer enclosure.
Containment Spray wash down volume between the outside of the Secondary Shield wall and the inside of the outer containment wall (referred to as Outside Missile Barrier, OMB), will not be obstructed in the path to the emergency recirculation sump. Two access openings exist at containment elevation 377 ft to control access to IMB from OMB. The size of each opening is a nominal 3 ft wide x 7 ft high; a screen door (1-1/2 inch diamond mesh) is installed at each location. This door is locked during normal plant operation to prevent personnel from entering the high radiation field in the IMB area.
Blockage of the openings to IMB is not likely because high-energy jets from the RCS will not cause direct debris in the OMB area. The Feedwater Line Break and the Main Steam Line Break accidents are the design basis accidents that may result in high-energy jets in the OMB area of containment. The ECCS response sequence to these accidents is not likely to progress further than the injection phase with the RWST as the water source. In fact, Emergency Recirculation from the containment emergency recirculation sumps is not modeled for these accidents.
Minimum containment flood level analyses have accounted for containment spray volumes that are trapped in the Refueling Cavity (Reference 8). The resulting flood levels do not degrade the post-LOCA recirculation function.
Blockage of containment drainage paths into the ECCS recirculation sumps is not a concern because there are no pipes that drain into the sumps.
1.1.1.6 The trash rack is a structure made of tube steel and angle steel supports, with side vertical stainless steel grating and a solid checkered plate cover. The top of the trash rack structure is approximately four (4) ft above the containment floor elevation of 377 ft. Design loads for the trash rack have been determined to address hydrostatic pressure, drag force, and debris impact.
Dynamic effects due to design basis high energy line breaks need not be considered in the structural qualification of the trash rack, and dynamic effects from pipe breaks considered in the vicinity of the trash rack structure need not be considered. An B/B - UFSAR A1.82-4 REVISION 12 - DECEMBER 2008 additional load due to lead shielding has also been considered to address lead shielding activities during refueling outages.
Design hydrodynamic loads includes the effects of sloshing. (Reference 1) 1.1.1.7 The recirculation sump screens at Braidwood and Byron Station are installed inside sump pits below the lowest containment floor elevation of 377 ft. The screens have been shown to be fully submerged following an accident. Air entrapment at the trash rack structure does not have any effect on the hydraulics of the sump screens.
1.1.1.8 All components of the trash rack structure are acceptable for the design basis loads consisting of hydrostatic pressure, drag force, debris impact loads, temporary lead shielding (For outage periods), rigging loads, pipe support loads, hydrodynamic loads, and seismic loading under OBE and SSE seismic events.
1.1.1.9 The sump screens and grating are made of stainless steel while the trash rack support elements are made of carbon steel. These elements are located in a dry environment and will not degrade during normal operation or for the duration of the accident mission time for the ECCS sump.
1.1.1.10 The debris interceptor structure (trash Rack) includes access openings to facilitate inspection of the trash rack, the sump screens and the sump outlets. The sump screens have been verified not to be susceptible to vortex formation during vendor testing.
1.1.1.11 The replacement sump screens have been verified by analysis and testing to result in a head loss that is adequate to maintain adequate NPSH for the RHR and CS pumps during the post-LOCA operation of the ECCS (Reference 2).
1.1.1.12 The possibility of debris-clogging flow restrictions downstream of the sump screen has been assessed to ensure adequate long term recirculation cooling, containment cooling, and containment pressure control capabilities. The downstream blockage evaluation has been performed in accordance with industry (Westinghouse Owners Group) and regulatory guidance, considering the size of the openings in the sump debris screen (1/12 inch).
B/B - UFSAR A1.82-5 REVISION 12 - DECEMBER 2008 (References 4, 5, and 6)
The downstream blockage evaluation has concluded that the Safety Injection throttle valves (_SI8810A-D, _SI8816A-D, _SI8822A-D) are susceptible to blockage.
The Safety Injection throttle valves have been modified to minimize the blockage potential after an accident. The modification installed a Copes-Vulcan HUSH II trim. This trim consists of an assembly of nested concentric cylinders each having a series of radially drilled holes. The orifice areas are developed by arranging the cylinders, one within the other, in an offset manner so that a series of restriction (pinch areas) and expansion areas occur in series. The total pressure is thus reduced in stages. The internal design dimensions of the trim assemblies have been set to minimize blockage due to debris that passes through the containment recirculation sump screens (hole size of 1/12 inch or 0.083 inch).
The final design of the valves' internal components has been developed based on results from extensive testing at Wyle Laboratories with debris-laden fluid. The quantity of debris was based on post-LOCA debris calculations specific to Byron and Braidwood Stations.
Testing results show a limited reduction in the valves Cv when coating debris is first added to the water mix. Testing also showed an increasing long term valve Cv after the full debris mix is added. For conservatism, the initial recorded Cv reduction is used in the hydraulic analysis (Reference 3) to calculate flow rates to the RCS for ECCS Cold Leg and Hot leg Recirculation under debris-laden conditions. The impact of the resulting flow rates on the accident analysis has been evaluated by Westinghouse and has been found to be acceptable (Reference 12).
B/B - UFSAR A1.82-6 REVISION 12 - DECEMBER 2008 1.1.1.13 The pump suction inlets at the sump screens have been verified by testing not to be susceptible to air ingestion or any other adverse hydraulic effect (Reference 7).
1.1.1.14 The Byron and Braidwood Station design does not include any drain pipe that would bypass the sump screens.
1.1.1.15 The design of the Byron and Braidwood sump screens, combined with the limited amount of fibrous materials inside the containment building, prevents the formation of a thin bed on the screen surfaces. This fact has been demonstrated during testing. 1.1.2 Minimizing Debris Byron and Braidwood Stations have minimized the debris that could reach the sumps by replacing fibrous insulation within its specific Zone of Influence (ZOI) on the Steam Generators and connected piping for Byron Unit 1 and Braidwood Unit 1. Trash rack gratings has been installed at elevation 377 ft of B/B - UFSAR A1.82-7 REVISION 12 - DECEMBER 2008 containment to minimize the quantity of debris that enters the sumps. 1.1.2.1 Braidwood and Byron have implemented a "Containment Loose Debris Inspection" procedure (References 15 and 16). The procedures outline the steps necessary to verify the containment is free of loose debris. It is applicable for all accessible areas just prior to establishing Containment Integrity and for affected areas at the completion of any Containment entry when Containment Integrity is already set. These procedures for containment closeout necessitates that a containment walkdown be performed for housekeeping deficiencies. The procedure incorporates a list of all unresolved housekeeping and equipment discrepancies and requires that resolution be included in the plant restart documentation. The procedure also provides guidance on general cleanliness and debris inspection guidelines.
1.1.2.2 Fibrous Insulation (Thermal Wrap Trademark) on the Steam Generators and connected piping has been replaced within its Zone of Influence (ZOI) for Byron Unit 1 and Braidwood Unit 1.
Procedures already exist to clean up the work area following maintenance activities inside containment. This action prevents the generation of additional latent debris.
1.1.2.3 Bare metal surfaces inside containment that are not stainless steel are coated. Galvanized steel surfaces (i.e., scaffolds) have been accounted for in the chemical effects evaluation for sizing the replacement screens.
1.1.3 Instrumentation Byron and Braidwood Stations do not rely on operator actions to mitigate the consequences of the accumulation of debris on the ECC sump screens.
1.1.4 Active Sump Screen System Byron and Braidwood Station do not employ an active screen.
B/B - UFSAR A1.82-8 REVISION 12 - DECEMBER 2008 1.1.5 Inservice Inspection To ensure the operability and structural integrity of the trash racks and screens, access openings can be used to inspect the ECC sump structures and the sump suction pipes. The sump screens, suction pipe, and trash racks are inspected each refueling outage in accordance with the requirements of the Technical Specifications.
1.2 Evaluation of Alte rnative Water Sources Byron and Braidwood Stations do not rely on operator actions to prevent the accumulation of debris on the sump screens.
1.3 Evaluation of Long-Term Recirculation Capability The nuclear industry has developed a methodology (Document #NEI 04-07) to perform the required evaluations with the cooperation of the Nuclear Energy Institute (NEI) and the NRC has issued a Safety Evaluation Report on the NEI document in December of 2005. The most important elements of the NEI methodology are break selection, debris generation, and debris transport. Substantial conservatism is built into the evaluation process. For example, although application of the Leak-before-Break methodology has been approved for Byron and Braidwood on the main Reactor Coolant Loop piping, the debris generation analysis assumes the largest pipes in the Reactor Coolant System break. Byron and Braidwood have followed the NEI methodology. The results of the evaluation concluded that the existing screens must be replaced.
New screens have been designed and manufactured by Control Components Inc. (CCI).
In addition, chemical reactions between the post-LOCA water and materials (Aluminum, copper, concrete) located in the containment building may result in chemical precipitants forming when the post-LOCA water temperature decreases in the latter part of the accident. This is called "chemical effects"; the design of the replacement screens is required to incorporate the impact on head loss due to "chemical effects".
Head loss testing on a scaled version of the replacement screens has been performed using scaled quantities of debris (coating, fiber insulation, reflective metal insulation, glass). The replacement screen assemblies have been verified to have a head loss that maintains adequate margin for the Net Positive Suction Head (NPSH) for the RHR and CS pumps when they take suction from B/B - UFSAR A1.82-9 REVISION 12 - DECEMBER 2008 the containment recirculation sumps. The head loss testing performed by CCI also includes chemical effects.
1.3.1 Net Positive Suction Head of ECCS and Containment Heat Removal Pumps 1.3.1.1 The NPSH analysis (Reference 2) for temperatures above 200 °F assumes that the vapor pressure of the recirculation sump liquid is equal to the containment pressure. This ensures that credit is not taken for increase in the containment pressure due to the accident.
The NPSH analysis for temperatures below 200 °F credits the minimum containment air pressure that was present inside containment before the accident. No credit is taken in the NPSH analysis for increase in containment pressure due to the accident.
B/B - UFSAR A1.82-10 REVISION 15 - DECEMBER 2014 Below is the final NPSH margin (NPSH Available minus Screen Head Loss minus NPSH Required) considering air evolution effects:
Tsump (°F) Screen Head Loss (ft)
RHR Pumps NPSH Margin (ft) Braidwood CS Pumps NPSH Margin(ft)
Byron CS Pumps NPSH Margin (ft) 258.1 4.13 2.77 0.67 1.07 250.6 4.13 2.27 0.57 0.97 242.5 4.13 2.17 0.57 0.97 230.0 4.13 2.67 0.67 1.07 210.0 4.20 2.7 0.60 1.0 203.9 4.20 2.7 0.60 1.0 200.0 4.20 2.8 0.60 1.0 195.0 4.24 6.66 4.46 4.96 175.0 4.55 13.35 10.95 11.45 155.0 4.77 17.63 15.23 15.63 135.0 5.34 19.46 16.96 17.46 120.001 6.11 18.89 16.59 16.99 119.999 6.11 17.19 15.19 15.59 95.0 7.20 17.0 14.9 15.3 73.4 8.27 15.73 13.73 14.13 Additional limitations exist on the allowable head loss for the recirculation sump screens due to the screen structural limit and the pump air void fraction limits. The resulting screen head loss margins are as follows:
Tsump (°F) Screen Head Loss (ft) Margin Over Screen Loss:
RHR Pumps (ft) Margin Over Screen Loss:
Braidwood CS Pumps (ft)
Margin Over Screen Loss:
Byron CS Pumps (ft) 258.1 4.13 2.8 0.7 1.1 250.6 4.13 2.3 0.6 0.9 242.5 4.13 2.2 0.6 0.9 230.0 4.13 2.6 0.7 1.0 210.0 4.20 2.7 0.6 1.0 203.9 4.20 2.7 0.6 1.0 200.0 4.20 2.8 0.6 1.0 195.0 4.24 6.7 4.5 4.9 175.0 4.55 11.0 11.0 11.0 155.0 4.77 10.8 10.8 10.8 135.0 5.34 10.2 10.2 10.2 120.001 6.11 9.4 9.4 9.4 119.999 6.11 9.4 9.4 9.4 95.0 7.20 8.3 8.3 8.3 73.4 8.27 7.3 7.3 7.3 B/B - UFSAR A1.82-11 REVISION 12 - DECEMBER 2008 Notes:
- The screen vendor reports the screen head loss calculation at specific temperatures up to a maximum of 212 °F. For temperatures greater than 212 °F, the screen head loss is conservatively assumed to be 4.13 feet, the value for expected head loss at 212 °F. No credit is taken for lower head loss due to lower viscosity at higher temperatures.
- The suction piping friction losses that have been used in the calculation of available NPSH have been determined based on maximum flow rates that are larger than calculated. This assumption accounts for approximately 1 ft of margin.
- The suction piping friction losses that have been used in the calculation of available NPSH have been determined based on the maximum calculated increase in viscosity due to chemical effects. The maximum viscosity increase does not apply to temperatures above 140 °F.
- 1.3.1.2 The NPSH analysis does not credit increase in the containment pressure due to the accident.
1.3.1.3 The NPSH analysis for the CS a nd RH pumps does not take credit for operation of the pum ps while cavitating.
For the SI and CV pumps, the NPSH analysis from the RWST is more limiting.
1.3.1.4 The post-accident temperature hi story for the containment and recirculation sump water has b een taken from the existing analyses for containment int egrity (Reference 2).
1.3.1.5 The hot channel correcti on factor specif ied in ANSI/HI 1.1-1.5-1994 is not used in de termining the margin b etween the available and required NPSH fo r ECCS and containme nt heat remo val system pumps. 1.3.1.6 The calculation of ava ilable NPSH uses t he containment flood level from the minim um containment flo od level calculation (Reference 8). This i nput minimizes the height of water above the pump suction (i.e., the level of water on the containment floor). The calculated m inimum flood height inside containment does not consider quantities of water that do not contribute to the sump pool (e.g., atmospheric steam, inactive water volume, B/B - UFSAR A1.82-12 REVISION 12 -
DECEMBER 2008 pooled on the containment floor and in t he refueling canal, spray droplets and other fal ling water, etc.).
1.3.1.7 The calculation of pipe and fitting resistan ce has been done using the Flowseries s oftware (Reference 3). The nominal screen resistance without b lockage by debris has been measured during model testing at the scr een vendor facility.
1.3.1.8 Head loss through the sump scr eens has been dete rmined based on testing of a model screen asse mbly, under desi gn basis debris loading conditions (sc aled accordingly).
1.3.1.9 The Calculation of ava ilable NPSH has be en performed as a limiting analysis that is applicable for the duration of the LOCA event. 1.3.2 Debris Sources and Generation 1.3.2.1 A number of breaks in each hig h-pressure system that relies on recirculation are cons idered to ensu re that the breaks that bound variations in debris g eneration by the size, quantity and type of debris are identifie d (Reference 9).
Based on various postulated break lo cations, the following break locations were evaluat ed per the methodology in the guidance document NEI 04-07, as modified by t he NRC's Saf ety Evaluation Report, to maximize the postulated debris created:
- 1. The interim leg at the inlet to the loop D reactor coolant pump (RCP) at approximate elevation 386'-0" is the largest postulated line in containment and will affect a large amount of fiber (Transco RMI and reflective mirror insulation on the major equipment and piping inside the missile barrier). It also is the most direct path to the sump. This is the limiting break for Braidwood and Byron since it has the greatest coating debris quantity, which dominates the fiber/pa rticulate head loss.
- 2. The loop A cold leg between the reactor coolant loop isolation valve and the reac tor shield wall at elevation 393'-0" is chosen because it is another large break that will create the greate st mix of insulati on debris types.
B/B - UFSAR A1.82-13 RE VISION 12 - DECEMBER 2008 It is also located farther from the sump, which will create a different transport path for debris.
- 3. The loop D hot leg bet ween the valve and the reactor shield wall at elevation 39 3'-0" is chosen be cause it generates the largest amount of fi ber debris in Unit 1 of Braidwood and Byron. 4. Additionally, an alternate break is evaluated at the 14-inch pressurizer surge line (branch off of the reactor coolant system loop D line) at t he connection to the pressurizer.
This break would damage the reflective mirro r insulation on most piping in loop D and a small amount in loop A, with the exception of piping near the top of the pressurizer. The loop D RCP and pressurizer insulation would also be damaged.
For Unit 1 of Braidwood and Byron only, the fiber insulation on the loop D SG will also be damaged.
1.3.2.2 Insulation The insulation for most lines and equipment is nearly identical in all four units. The majority of the insulation is Mirror RMI. Sections for the SGs in Unit 1 for both Braidwood and Byron are insulated with Transco RMI and Transco Thermal Wrap.
The associated Braidwood and Byron Unit 1 SG piping connections (Main Steam, Feedwater, Auxiliary Feedwater, Steam Generator Blowdown) also have sections of Transco Thermal Wrap. The thermal wrap insulation that was located within the Zone of Influence (ZOI) for this insulation type has been replaced with reflective metal insulation for Braidwood Unit 1 and Byron Unit
- 1. The ZOI characteristics for the RMI and thermal wrap were applied using the criteria established in the NRC Generic Letter (GL) 2004-02 Safety Evaluation, Table 3-2.
Coating In accordance with the NRC's Safety Evaluation Report for NEI 04-07, a ZOI of ten pipe diameters (10D) was used for the qualified coating. All unqualified coating was assumed to fail regardless of location inside the containment.
Foreign Material The quantity and type of foreign material inside containment was based on walkdown data performed for both units at Byron and Braidwood. The foreign material included self-adhesive labels, stickers, placards, etc. The foreign material includes all identified foreign material in containment, and per the above B/B - UFSAR A1.82-14 REVISION 12
- DECEMBER 2008 referenced guidance, appropriate quantities were assumed to transport to the sump. In addition, 200 ft2 of degraded qualified coating was considered in the debris mix.
Latent Debris A latent debris walkdown was performed at Byron and Braidwood in accordance with the NRC's Safety Evaluation Report for NEI 04-07, Section 3.5. Using a masolin cloth, samples were collected from the various surfaces at different floor elevations and when practical, different locations on each floor. When a surface was not accessible for sampling, an alternate surface was selected and noted on the walkdown report, such as circular pipe for an inaccessible circular duct. The net weight differences between the pre-sample and post-sample weight were used to statistically extrapolate the amount of latent debris for the entire containment using a 90% confidence level.
1.3.2.6 In addition to debris generated by jet forces from the pipe rupture, debris created by the resulting containment environment (thermal and chemical) have been considered in the analyses.
Examples of this type of debris would be disbondment of coatings in the form of chips and particulates or formation of chemical debris (precipitants) caused by chemical reactions in the pool.
Head loss testing for the replacement screens has been performed using debris that included post-accident chemical effects.
1.3.2.7 All insulation within the ZOI has been accounted in the debris term. Continued degradation due to cascading water is irrelevant.
1.3.3 Debris Transport The transport of the debris from the break location to the sump screen is evaluated (Reference 10) using the methods outlined in section 3.6 of NEI 04-07 as amended by the NRC SER. The means of transport considered are blowdown, washdown, pool fill, and recirculation for all types of debris. The recirculation transport analysis was performed by Sargent & Lundy using computational fluid dynamics (CFD) models developed using the computer program FLUENT. The CFD models were created by RWDI, Inc. Outputs of the CFD analysis include global (entire containment) and local (near sump pit) velocity contours, turbulent kinetic energy (TKE) contours, path lines and flow distributions for various scenarios. All particulate and coating debris was modeled as fines and 100% transports to the B/B - UFSAR A1.82-15 REVISION 12 -
DECEMBER 2008 screen. The debris transport phenomena due to the blowdown, washdown, pool fill-up, and recirculation transport modes are summarized using debris transport logic trees consistent with the Drywell Debris Transport Study (DDTS) documented in NUREG/CR-6369, "Drywell Debris Transport Study." The debris transport logic trees consider the effect of dislocation, hold up on the floor or other structures, deposition in active or inactive pools, lift over curbs, and erosion of debris.
Miscellaneous (foreign material) debris (tape, labels, etc.) is included in the debris load, and considered in the screen design as a sacrificial area. All miscellaneous debris is assumed to be 100% transportable.
The following is a summary of the overall transport fractions for all debris types:
Debris Type Overall Transport Fraction RMI 0.142 Qualified Coatings 1.00 Unqualified Coatings 1.00 Latent Debris 1.00 Foreign Material 1.00 The transport fractions presented above are bounding for all break locations, including a break in the RCS piping above the sump. 1.3.3.1 The debris quantities that have been used in the design basis testing for the filters are bounding values. No credit is taken for reduced debris accumulation at the filters that may be due to the actual sequence of debris accumulation at the filters during the event.
1.3.3.2 Based on the results of the CFD analysis, all (100%) coating, latent, and foreign material debris transports to the sump screen for all scenarios, except for Reflective Metal Insulation Debris.
B/B - UFSAR A1.82-16 REVISION 12 -
DECEMBER 2008 NUREG/CR-3616 documents debris transport properties for stainless steel RMI. The following observations pertaining to RMI transport were made.
- Thick sheets of foil require higher velo cities for transport than thin sheets; i.e.
transport velocity tends to increase with mat erial thickness
- Crumpled foil tends to transport at lower velocities than uncrumpled foil
- Velocity of motion of samples (crumpled or u ncrumpled foil) is much less than the flow velocity
- RMI transport modes include fold ing, tumbling, rolling, and sliding along the floor
- RMI does not become "waterborne" during transport; i.e. a portion of the foil is always in contact with the floor.
Therefore the velocity contours at 1 inch above the floor are considered accep table for RMI
- Walls tend to hi nder transport
- Interaction of foil pi eces with each o ther often causes jamming and immobilization of the pieces; high flow velocities are then required to break up jams and resume transport
- Because RMI does not become "w aterborne" during transport; i.e. a portion of th e foil is always in contact with the floor, it does not cause screen blockage to a height greater than the height and width of the debris; i.e. the RMI accumulates on the f loor when it tra nsports to a screen Results from the Computational Fluid Dynamic Analysis of containment show high velocity (>0.40 ft/s) transport paths to the sump pit area for all scenarios modeled. Therefore, all RMI debris can be transported to the vicinity of the sump, regardless of whether 1 or 2 trains are operating. However, the trash rack debris retainer has been designed to prevent large RMI debris (debris too large to pass through a 4-inch by 4-inch opening) from being transported to the sump. Per NUREG/CR-3616 (Ref.7.4.1) RMI debris transports by rolling, tumbling, and sliding and does not become "waterborne" (see §4.1.5.4). Since the large RMI debris would have to become "waterborne" to transport over the trash rack's approximately fourteen inch long by approximately six inch high debris retainer, it will be retained and accumulate on the floor in front of the trash rack.
With conservatively equating the RMI debris larger than six inches with the debris too large to pass through a 4-inch by 4-inch opening only the RMI debris less than 6 inches transports B/B - UFSAR A1.82-17 REVISION 12 -
DECEMBER 2008 to the sump screens. Per Reference 11, a debris pile may form in front of the trash rack debris retainer. Large RMI debris may be drawn through the trash rack openings by climbing over this pile. This mode of transport would be restricted by the interaction of the foil pieces which often causes jamming and immobilization, the height of the debris pile, and the presence of the horizontal debris retainer. The conservatism included in the calculation of the amount of small debris that is transported to the sump screen will bound the minor fraction of the large debris that may transport past the trash rack by this path.
B/B - UFSAR A1.82-18 REVISION 12 -
DECEMBER 2008 The figure below gives the Transport Logic Tree for RMI debris:
BlowdownWashdownPoolFill-upRecirculationTrash RackPathFractionDeposition LocationStalled0.0010.00Not transportedActive poolRetained by Rack1.000.858Transport20.858Not transportedContainment Floor1.00Transport1.000.14230.142Sump screenInactive pool 0.0040.00Not transported0.858Not Transported0.142Sump Screen RMI transport fractions that were determined using the logic trees are summarized in the table below:
Size Debris Transport Fraction Fraction of Debris at Sump Screen Debris Distribution (Applicable to All Scenarios) (Applicable to All Scenarios)
Type (Fraction)
Before CS After CS Before CS After CS Mirror RMI
< 2 inches 0.061 1.00 1.00 0.061 0.061 >2 inches, < 6 inches 0.081 1.00 1.00 0.081 0.081 > 6 inches 0.858 0.0 0.0 0.0 0.0 Sum 1.0 - - 0.142 0.142 Thus it can be seen that 14.2% of Mirror RMI foil debris transports to the sump screen for all scenarios.
Large RMI debris, relatively intact RMI, end covers, etc. due to RCS line breaks above the sump do not transport to the sump screen because the screen is sufficiently protected from blowdown debris by the top plate of the trash rack.
1.3.3.3 The recirculation analysis (Reference 10) considers the maximum recirculation flow rates.
B/B - UFSAR A1.82-19 RE VISION 12 - DECEMBER 2008 The following is a summary of the overall transport fractions for all debris types:
Debris Type Overall Transport Fraction RMI 0.142 Qualified Coatings 1.00 Unqualified Coatings 1.00 Latent Debris 1.00 Foreign Material 1.00 The transport fractions presented above are bounding for all break locations, including a break in the RCS piping above the sump. 1.3.3.4 The debris transport analysis used computational fluid dynamics (CFD) simulations in combination with the experimental debris transport data.
The following is a summary of the overall transport fractions for all debris types:
Debris Type Overall Transport Fraction RMI 0.142 Qualified Coatings 1.00 Unqualified Coatings 1.00 Latent Debris 1.00 Foreign Material 1.00 The transport fractions presented above are bounding for all break locations, including a break in the RCS piping above the sump. 1.3.3.5 Both ECCS sump openings are protected by a trash rack above elevation 377 ft. The trash rack prevents heavier debris from entering the sump pits.
1.3.3.6 All debris that has been evaluated to reach the sump pits is accounted for in the head loss analysis.
1.3.3.7 The head loss evaluation assumes that the 100% debris load is present at the sump filters at the time the RHR pump suction switches over to the Recirculation sumps. Also, maximum flow rates are considered, including the flow from the Containment Spray pump.
B/B - UFSAR A1.82-20 RE VISION 12 - DECEMBER 2008 1.3.3.8 The debris quantity that has been calculated to reach the sump screens does not apply any reduction factor due to the results of airborne or containment spray washdown debris transport analyses.
Flood level analyses inside containment show that the minimum flood level will be sufficient to fully submerge the recirculation sump filters. The minimum flood level analysis accounts for water inventory hold-up inside containment.
1.3.3.9 The effects of floating or buoyant debris have been evaluated for the integrity of the trash rack. The effects of floating or buoyant debris on head loss are not evaluated because the sump filters are fully submerged.
1.3.4 Debris Accumulat ion and Head Loss 1.3.4.1 The debris accumulation on the sump filters accounts for the total debris quantity that has been calculated. The full debris quantity (scaled accordingly) has been used in the vendor head loss testing.
1.3.4.2 The containment recirculation sump screens are entirely located within the sump pit below the containment floor elevation of 377 ft. Calculations for minimum containment flood level have demonstrated that the sump screens will be submerged at the time of ECCS Switchover to the containment recirculation sump. The CS pumps are switched over to the containment recirculation sumps later than the RHR pumps. The flood level increases as the LOCA event progresses; thus, the sump screens will be submerged at the time of CS Switchover.
The head loss through the sump screens has been determined by testing a model of the screens under debris loading conditions scaled accordingly from the debris quantities that have been calculated specific to Byron and Braidwood.
1.3.4.3 The NPSH analysis at high temperature follows the current Byron and Braidwood licensing basis. The containment pressure and sump water vapor pressure are assumed to be equal. This assumption assures that credit is not taken for increases in containment pressure due to the accident.
B/B - UFSAR A1.82-21 RE VISION 12 - DECEMBER 2008 As part of the chemical effects evaluations related to head loss through the containment recirculation sump strainers, the NPSH analysis for the RHR pumps has been performed at low temperatures. In accordance with the requirements specified in Regulatory Guide 1.1, the NPSH analysis at low temperatures assumes the containment atmospheric pressure is equal to the minimum containment atmospheric pressure that would be present inside containment before the Loss of Coolant Accident (LOCA) event. This analysis does not credit calculated increases in containment pressure as a result of the LOCA.
1.3.4.4 The sump screens at Byron and Braidwood are fully submerged at the times the RH and CS pumps' suctions are switched over to the containment recirculation sumps.
1.3.4.5 The head loss through the recirculation sump screens has been determined by testing. The head loss through the recirculation screen assembly, downstream of the screens, has been determined via a Computational Fluid Dynamic (CFD) analysis. (Reference 7).
The total head loss through the recirculation sump screens has been determined to be 4.20 ft at a temperature of 200 °F.
1.3.4.6 The screen head loss has been determined by testing based on limiting debris quantities specific to Byron and Braidwood.
B/B - UFSAR A1.82-22 RE VISION 12 - DECEMBER 2008
References:
- 1. Design Analysis #6.1.2.6-BRW-06-0029-S (Braidw ood),
- 6.1.2.6-BYR06-031 (Byron)
- 2. Design Analysis #BRW-06-0035-M (Braidwood), #BYR06-058 (Byron) 3. Design Analysis #BRW 0016-M (Braidw ood), # BYR06-029 (Byron) 4. Design Analysis #BRW-05-0061-M (Braidwood), #BYR05-043 (Byron) 5. Design Analysis #BRW-05-0063-M (Braidwood), #BYR05-061 (Byron) 6. Design Analysis #BRW-05-0084-M (Braidwood), #BYR06-017 (Byron) 7. Design Analysis
- 3 SA-096.018
- 8. Design Analysis #SI-90-01
- 9. Design Analysis #BRW-05-0059-M (Braidwood), #BYR05-041 (Byron) 10. Design Analysis #BRW-05-0060-M (Braidwood), #BYR05-042 (Byron) 11. Design Analysis #BRW-06-0015-M (Braidwood), #BYR06-025 (Byron) 12. Design Analysis
- CAE-07-49/CCE-07-48
- 13. Letter from K. R. Jury (Exelon Generation Company, LLC) to U. S. Nuclear Regulatory Commi ssion, "Supplement to Exelon Response to NRC Generic Letter 2004-02, 'Pot ential Impact of Debris Blockage on Emerge ncy Recirculation During Design Basis Accidents at Pressurized-Water R eactors,' " dated May 31, 2006
- 14. Letter from R. F. Kuntz (U. S. Nuclear Regulatory Commission) to C. M. Crane (Exelon Generation Company, LLC), "Byron Station, Unit N o.1 and Braidwoo d Station, Unit 2 - Requested Extens ion of Completion Schedule for NRC Generic Letter 2004-02, "Potential Imp act of Debris Blockage on Emergency Recircul ation During D esign-Basis Accidents at Pressur ized-Water Reactors,'
" dated July 21, 2006 15. Procedure BwOS TRM 2
.5.b.1 for Braidwood
- 16. Procedure BOSR Z.5.b.1-1 for Byron B/B-UFSAR A1.83-1 REVISION 15 - DECEMBER 2014 REGULATORY GUIDE 1.83 INSERVICE INSPECTION OF PRESSURIZED WATER REACTOR STEAM GENERATOR TUBES Regulatory Guide 1.83 de scribes an acceptable method of complying with the Commission's regulati ons with regard to inservice inspection of pressurized water reactor steam ge nerator tubes.
The plant design includes the fe atures of Regula tory Position C.1.
The preservice and inser vice inspection of steam generator tubing was conducted in accorda nce with Regulatory Posi tions C.2 through C.8 of Regulatory Guide 1.83, Revision 1, as modified by the Technical Specifications for Byr on/Braidwood. N uclear Energy Institute (NEI) 97-06 "Steam Generator P rogram Guidelines,"
supercedes NRC Regulatory Guide 1.83 for ins ervice inspection requirements. NEI 97-06 was approved for use by the NRC via License Amendments 150 a nd 179 for Byron and License Amendments 144 and 172 for Braidwood.
B/B-UFSAR A1.84-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.84 DESIGN AND FABRI CATION CODE CASE ACCEPTABILITY ASME SECTION III DIVISION 1 The Licensee complies with the regulatory position. This regulatory guide lists t hose section III ASME code cases relevant to design and fabrication that are generally acceptable to the NRC for implementation in the licensing of light-water-cooled nuclear power plants.
Code cases expla in the intent of cod e rules or provide alternative requirements und er special c onditions.
Implementation of indi vidual code cases is limited to the requirements as specified in the inquiry and reply sections of each code case. The ASME considers the use of code cases to be optional for the user and not a mandatory requirement. Use of this regulatory guide is optional.
Approval of code cases l isted in this regulatory guide is by code case number and date of ASME approval. Compon ents ordered to a specific version of a code case need not be changed because a subsequent revision to the code case is listed as the approved version in subsequent revisions of the r egulatory guide.
Similarly, components or dered to a code case that was previously approved for use need not be cha nged because the code case has been subsequently annulled.
Code cases on the approved list may be applied to co mponents that were in the process of c onstruction prior to the effective date of the code case within the limits specified in the code case and applicable regulations, or recommended in other regulatory guides.
Code cases listed in this re gulatory guide are generically acceptable for i mplementation. Beginning with Revision 16 to this regulatory guide, it is no longer n ecessary to obtain NRC approval to use code c ases listed in the regulatory guide.
Code cases not l isted in this regulato ry guide cannot be implemented unless formal ap proval is ob tained from the Commission in accordance with footnote 6 of the Codes and Standards Rule, 10CFR 50.55a.
Components with long lead times were o rdered prior to the original effective dates for R egulatory Guides 1
.84 and 1.85.
Nevertheless, there are no known examples of code cases being applied to components, except those approved by either Regulatory Guide 1.84 or 1.85, with the fol lowing exceptions or special conditions:
B/B-UFSAR A1.84-2 REVISION 6 - DECEMBER 1996 Code Case 1528: Fracture toughne ss information f or this code used in the construc tion of the steam generators and pressuriz ers has been supplied to the NRC by WCAP-929 2, March 1978, "Dynamic Fracture Toughness of ASME SA-508 Class 2a and ASME SA-533 Grade A Cl ass 2 Base and Heat Affected Zone Material a nd Applicable Weld Metals."
Code Case 1637: This code ca se was used for the purchase of heat exchanger t ubing. Authoriz ation for its use was obtained fro m the NRC.
Refer to Subsections 5.4.2 and 5.2.1 for further information.
In addition, the following code cases have been approved by the NRC. This list is s ubject to change bas ed on regulatory guide revisions. Code cases approved for use by R egulatory Guide 1.84 need not be listed here prior to being i mplemented.
REGULATORY A SME CODE GUIDE A PPROVAL CASE REVISION DATE TITLE N-272 18 05/15/80Compiling Data Repor t Records, Section III, Division 1 N-275 18 05/18/80Repair of Welds, Section III, Division 1 N-292 19 01/05/81Depositi ng Weld Metal Prior to Preparing Ends for Welding, Section III, Div ision 1, Class 1, 2, and 3 N-340 N/A 06/17/82 A lternate Rules for Examination of Weld Edge Pre paration, Section
III, Division 1, Classes 1, 2, MC, and CS. (Licens ee to Provide Justification Each T ime Code Case N-340 is Used.) N-403 N/A 02/14/85Reassembly of Subsection NF Component and Pi ping Supports, Section III, Division 1 N-411 24 09/17/84 A lternative Damp ing Values for Seismic Analysis of Class 1, 2, and 3 Piping, Section III, Division 1 N-413 24 02/14/85Minimum Size of Fillet Welds for Subsection NF Linear Type
Supports, Section III, Division 1
B/B-UFSAR A1.84-3 REVISION 3 - DECEMBER 1991 The status of code case approval is co ntinually changing; however, the rules for use of th is regulatory gu ide normally do not change. Therefore, the above discussions are applicable to any revision of this regulatory guide, provi ded the limitations of the regulatory guide revision are adhered to.
B/B-UFSAR A1.85-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.85 MATERIALS CODE ACCEPTABILITY ASME SECTION III DIVISION 1 The Licensee complies with the regulatory position. This regulatory guide lists t hose Section III ASME code cases relevant
to materials and testing that are generally acceptab le to the NRC for implementation in the licens ing of light-water-c ooled nuclear power plants.
Code cases expla in the intent of Cod e rules or provide alternative requirements und er special c onditions.
Implementation of indi vidual code cases is limited to the requirements as specified in the inquiry and reply sections of each code case. The ASME considers the use of code cases to be optional for the user and not a mandatory requirement. Use of this regulatory guide is optional.
Approval of code cases l isted in this regulatory guide is by code case number and date of ASME approval. Compon ents ordered to a specific version of a code case need not be changed because a subsequent revision to the code case is listed as the approved version in subsequent revisions of the r egulatory guide.
Similarly, components or dered to a code case that was previously approved for use need not be cha nged because the code case has been subsequently annulled.
Code cases on the approved list may be applied to co mponents that were in the process of c onstruction prior to the effective date of the code case within the limits specified in the code case and applicable regulations, or recommended in other regulatory guides.
Code cases listed in this re gulatory guide are generically acceptable for i mplementation. Beginning with Revision 16 to this regulatory guide, it is no longer n ecessary to obtain NRC approval to use code c ases listed in the regulatory guide.
Code cases not l isted in this regulato ry guide cannot be implemented unless formal ap proval is ob tained from the Commission in accordance with footnote 6 of the Codes and Standards Rule, 10CFR 50.55a.
Components with long lead times were o rdered prior to the original effective dates for R egulatory Guides 1
.84 and 1.85.
Nevertheless, there are no known examples of code cases being applied to components except those appro ved by either Regulatory Guide 1.84 or 1.85, with the fol lowing exceptions or special conditions:
B/B-UFSAR A1.85-2 REVISION 3 - DECEMBER 1991 Code Case 1528: Fracture toughne ss information f or this code used in the construc tion of the steam generators and pressuriz ers has been supplied to the NRC by WCAP-929 2, March 1978, "Dynamic Fracture Toughness of ASME SA-508 Class 2a and ASME SA-533 Grade A Cl ass 2 Base and Heat Affected Zone Material a nd Applicable Weld Metals."
Code Case 1637: This code ca se was used for the purchase of heat exchanger t ubing. Authoriz ation for its use was obtained fro m the NRC.
Code Case N-242-1: Paragraphs 1 t hrough 4 of th is code case are used in the acceptance of limited amounts of materials in ASME Se ction III systems.
Refer to Subsections 5.4.2 and 5.2.1 for further information.
In addition, the following code cases have been approved by the NRC. This list is s ubject to change bas ed on regulatory guide revisions. Code cases approved for use by R egulatory Guide 1.85 need not be listed here prior to being i mplemented.
REGULATORY A SME CODE GUIDE A PPROVAL CASE REVISION DATE TITLE N-242-1 18 04/10/80Mate rial Certification, Section III, Division 1, Classes 1, 2, 3, MC, and CS Construction N-295 19 01/15/81Use of Previously Produced Material, NC A-1140 The status of code case approval is co ntinually changing; however, the rules for use of th is regulatory gu ide normally do not change. Therefore, the above discussions are applicable to any revision of this regulatory guide, provi ded the limitations of the regulatory guide revision are adhered to.
B/B-UFSAR A1.86-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.86 TERMINATION OF OPERATI NG LICENSES FOR NUCLEAR REACTORS
Revision 0 of Regulatory Guide 1.86 will be complied with by the regulatory staff for the termination of operating licenses at the end of the station design life.
B/B-UFSAR A1.87-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.87 GUIDANCE FOR CONSTRUCTION OF C LASS 1 COMPONENTS IN ELEVATED-TEMPERATURE REACTORS This guide is not pertinent to this applicat ion, since the Byron/Braidwood reactors are not "high-temperature reactors."
B/B-UFSAR A1.88-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.88 COLLECTION, STORAGE, AND MAINTEN ANCE OF NUCLEAR POWER PLANT QUALITY ASSURANCE RECORDS The requirements of Regu latory Guide 1.88 ha ve been incorporated in Regulatory Guide 1.28, Revision 3. Regul atory Guide 1.88 was withdrawn on June 17, 1991.
The Licensee complies with the intent of Reg ulatory Guide 1.28 Revision 3, but applies it to ANSI/ASME NQA-1-1994.
B/B-UFSAR A1.89-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.89 QUALIFICATION OF CLASS lE EQUIPMENT FOR NUCLEAR POWER PLANTS
NSSS Scope For Westinghouse NSSS Cl ass lE equipment, We stinghouse has met the requirements of IEEE
-323-1974, "IEEE Stand ard for Qualifying Class lE Equipment f or Nuclear Power G enerating Stations" including the Nuclear Power Engineering Commit tee (NPEC) Position Statement of July 24, 1975, and Regulatory Gui de 1.89, Revision 0, by providing an app ropriate combination of the following:
type testing, operating experience, qualificat ion by analysis, and on-going qualificati on. This commitment has been satisfied by NRC review and ap proval of Westinghou se Topical Report WCAP-8587.
Non-NSSS Scope
The extent of the Licensee's c ommitment to com ply with the requirements of Regulato ry Guide 1.89 is presented in Subsections 3.11.2 and 8.1.16.
The Licensee follows the provisi ons in Revision 1 of Regulatory Guide 1.89 to qualify replacement electric equip ment installed subsequent to February 22, 1993. Revisi on 1 provides NRC-accepted reasons to allow ex ception to qualification requirements.
B/B-UFSAR A1.90-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.90 INSERVICE INSPECTION OF PRESTRESSED CONCRETE CONTAINMENT STRUCTURES W ITH GROUTED TENDONS The containment design d oes not use grouted te ndons. Thus, this guide is not applicable to Byron/Braidwood.
B/B-UFSAR A1.91-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.91 EVALUATION OF EXPLOSIONS POSTULATED TO OCCUR ON TRANSPORTATION ROUTE S NEAR NUCLEAR POWER PLANTS The plant design conforms to Rev ision 1 of this regulatory guide as described in Subs ection 2.2.3.2.
B/B-UFSAR A1.92-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.92 COMBINING MODAL RESPONSES AND SPATIAL COMPONENTS IN SEISMIC RES PONSE ANALYSIS The plant design conforms to Rev ision 1 of this regulatory guide as described in Subsecti ons 3.7.2 and 3.7.3.7.
B/B-UFSAR A1.93-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.93 AVAILABILITY OF ELEC TRIC POWER SOURCES
Availability of elec tric power sources is discussed in the Technical Specifications.
The Licensee complies wi th the requirements in R evision 0 of this guide with the follo wing exceptions and clarifications:
Regulatory Positions C.l, C.2, and C.4 refer to a 72-hour time interval for power operation when the available power sources are less tha n the "Limiting Conditions for Operation." The operating time limits delineated in regulatory positions C.1 through C.5 are explicitly for corrective maintenance a ctivities only. These operating time limits sh ould not be construed to include preventive m aintenance activit ies that require the incapacitation of any required electric power source. Therefore, per this guide, preventive maintenance should be schedu led for performance during cold shutdown and/or refueling periods.
The Licensee has determi ned that performance of
preventive maint enance activities on the system auxiliary transforme rs and Emergency Diesel Generators may be safely performed with both units at power.
Therefore, system auxili ary transformer and Emergency Diesel Generator pre ventive maintenance activities may be performed dur ing periods other th an cold shutdown and/or refueling.
B/B-UFSAR A1.94-1 REVISION 15 - DECEMBER 2014 REGULATORY GUIDE 1.94 QUALITY ASSURANCE REQU IREMENTS FOR INSTA LLATION, INSPECTION, AND TESTING OF STRUCTURAL CONC RETE AND STRUC TURAL STEEL DURING THE CONST RUCTION PHASE OF NUC LEAR POWER PLANTS
Regulatory Guide 1.94 endorsed ANSI N45.2.5, Quality Assurance Requirements for Installation, Inspection, a nd Testing of Structural Concrete and Structural Steel Dur ing the Construction Phase of Nuclear Power Plants. NQA-1-1994, Subpart 2.5, Quality Assurance Requirements for Installation, Ins pection, and Testing of Structural Concrete, Structural Steel, So ils, and Foundations for Nuclear Power Plants superse des this commitment to Regulatory Guide 1.94 and ANSI N45.2.5 as documented in the Quality Assurance Topical Report (NO-AA-10). For spec ific information relating to concrete standar ds, refer to Appendix B.
B/B-UFSAR A1.95-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.95 PROTECTION OF NUCLEA R POWER PLANT CONT ROL ROOM OPERATORS AGAINST AN ACCIDENTA L CHLORINE RELEASE The Licensee complies wi th the requirements in R evision 1 of this guide, as discussed in Subsection 6.4.4.
B/B-UFSAR A1.96-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.96 DESIGN OF MAIN S TEAM ISOLATION VALVE LEA KAGE CONTROL SYSTEMS FOR BOILING WATER REACTOR NUCLEAR POWER PLANTS The requirements of this guide are not applicable to pressurized water reactors.
B/B-UFSAR A1.97-1 REVISION 11 - DECEMBER 2006 REGULATORY GUIDE 1.97 INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR P OWER PLANTS TO ASSESS PLANT AND ENVIRONS CONDITIONS DURING AND FOLLOWING AN ACCIDENT
Compliance with Revision 3 of Re gulatory Guide 1.97 was discussed in two letters.
The Preliminary Rep ort on compliance to Regulatory Guide 1.97 is in a letter from K.A. Ainger of CECo to H.R. Denton of the NRC dated Februar y 27, 1987. The Final Report is in a letter from Steve Hunsader of CECo to T. E. Murley of the NRC dated September 1, 1 987. These transmitta ls, which furnished a report of compliance with Regu latory Guide 1
.97, met the license conditions for the Byron/Braidwood Stations. Refer to UFSAR Sections 6.2, 6.
5, 7.5, 11.5, E.21, E.30, and E.75 for further discussion.
The hydrogen monitoring system was originall y designed to meet the requirements of Regulato ry Guide 1.97 Cate gory 1 instruments.
Regulatory Guide 1.97 Category 1 is intended for key variables that most directly indicate the accomplishment of a safety function for design basis accident events an d provides f or full qualification, redundanc y, and continuous re al-time display and requires onsite (standby) power.
Based on a rev ision to 10 CFR 50.44 which eliminated t he design basis loss-of-coolant hydrogen release, the hydrogen monito rs have been reclassified as Regulatory Guide 1.97 Category 3 instruments.
B/B-UFSAR A1.98-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.98 ASSUMPTIONS USED FOR EVALUATING THE POTE NTIAL RADIOLOGICAL CONSEQUENCES OF A RADIOACTIVE OFF-GAS SYSTEM FAILURE IN A BOILING WATER REACTOR
The requirements of this guide are not applicable to pressurized water reactors.
B/B-UFSAR A1.99-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.99 RADIATION EMBRITTLEMENT OF R EACTOR VESSEL MATERIALS
Regulatory Guide 1.99 was issu ed after procurement of the reactor vessels.
The vessel materials u sed in the construction of the reactor vessels have been ev aluated using the me thods provid ed within Regulatory Guide 1.99, Revision 2. The end-of-life adjusted reference temperature (ART) at the 1/4 thick ness (1/4T) position in the vessel wall is less than 200
°F for each of the reactor vessels. The ART predictions, t herefore, are in agreement with Regulatory Position C.3.
Refer to UFSAR Secti on 5.3 for furth er discussion.
B/B-UFSAR A1.100-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.100 SEISMIC QUALIFICATION OF ELECTRIC EQUIPMENT FOR NUCLEAR POWER PLANTS NSSS Scope The Licensee is in com pliance with the objectives in Revision 1 of Regulatory Guide 1.10
- 0. The Westinghouse program for seismic qualification of safet y-related electrical equipment is delineated in WCAP-8587, Revision 1, which has b een reviewed by the NRC. For further detail s refer to S ection 3.10.
Non-NSSS Scope
The Licensee complies with the objectives in Revision 1 of Regulatory Guide 1.100.
The Licensee's approach to seismic qualification of Class lE equipm ent is discussed in Section 3.10.
Revision 1 of Regulatory Guide 1
.100 was in effect when the operating license appl ications were docketed.
B/B-UFSAR A1.101-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.101 EMERGENCY PLANNING AND PREPAREDNESS FOR NUCLEAR POWER REACTORS The guidance provided by Regulatory Guide 1.101, Revision 3, was
utilized in the preparat ion of the Licensee' s emergency response plans. The Licensee c omplies with this regulatory guide as described in the Exelon Nuclear Standardized Radiological Emergency Plan.
B/B-UFSAR A1.102-2 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.102 FLOOD PROTECTION FOR NUC LEAR POWER PLANTS
The plant design conforms to Rev ision 1 of this regulatory guide as described in Sect ions 2.4 and 3.4.
B/B-UFSAR A1.103-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.103 POST-TENSIONED PRESTRESSING SYSTEMS FOR CONCRETE REACTOR VESSELS AND CONTAINMENTS The plant design conforms to t he regulatory positions in Revision 1 as described in Su bsection 3.8.1 and Appendix B.3. The requirements remain in effect even though the regulatory guide was withdrawn on July 8, 1981. The regulatory positions are now covered by one or mo re national standards.
B/B-UFSAR A1.104-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.104 OVERHEAD CRANE HANDLING SYSTEMS FOR NUCLEAR POWER PLANTS
Regulatory Guide 1.104 w as withdrawn on Augu st 16, 1979 because the information it con tained was published in NUREG-0554, "Single-Failure-Proof Cr anes for Nuclear Pow er Plants." Byron and Braidwood Stations f ollow the recommendati ons of NUREG-0554.
B/B-UFSAR A1.105-1 REVISION 7 - DECEMBER 1998 REGULATORY GUIDE 1.105 INSTRUMENT SETPOINTS
The Licensee complies wi th the regulatory po sition in Revision 1 with the following exceptions keyed to paragraph numbers in the position. Revision 1 was in effect when the operating license applications were docketed.
Position C.5, requires l ocking devices on in strument setpoint adjustment mechanisms.
We have not spec ified such locking devices on our instrument data sheets, so that these would only be available to the ex tent that they are sta ndard equipment. In general, locking devices are not required to maintain stable instrument setpoints and we beli eve that setpoint stability will not be improved by provi ding locking devices.
Position C.6, requires d ocumentation of the as sumptions used in selecting setpoint values and the margins between the setpoints and the limiting safety system values. The docu mentation is to include definition of instrument setpoint dr ift rate and the relationship of the dr ift rate to testing intervals. The Byron/Braidwood design conforms to this position only to the degree that setpoints are documented on the instrument data sheets along with inst rument range and the maximum range of the parameter being measured.
With respect to t he other requirements of Position C.6, gener ic drift rates are not generally available for any instruments since drift rates would be affected by the particular service to which the instrument was subjected.
Testing intervals are set on the basis of past experience with the specific instrument types in question.
B/B-UFSAR A1.106-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.106 THERMAL OVERLOAD PROTECT ION FOR ELECTRIC MOTORS ON MOTOR-OPERATED VALVES The Licensee complies wi th the requirements in R evision 1 of this regulatory guide. T he Licensee has selected the method described in Regulatory Position C.2.
B/B-UFSAR A1.107-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.107 QUALIFICATIONS F OR CEMENT GROUTING FOR PRESTRESSING TENDONS IN CONTAINMENT STRUCTURES The Byron/Braidwood containment design does not use grouted tendons; therefore, this guide is not applicable to Byron and Braidwood stations.
B/B-UFSAR A1.108-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.108 PERIODIC TESTING OF DIES EL GENERATORS USED AS ONSITE ELECTRIC POWE R SYSTEMS AT NUCLEAR POWER PLANTS The guidance in Regula tory Guide 1.108 h as been updated and incorporated into Revision 3 of Regulatory Guide 1.9. Regulatory Guide 1.108 was withdrawn on July 19, 1993.
B/B-UFSAR A1.109-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.109 CALCULATION OF ANNUA L DOSES TO MAN FROM ROUTINE RELEASES OF REACTOR EFFLUENT S FOR THE PURPOSE OF EVALUATING COMPLIANCE WITH 10 CFR PART 50, APPENDIX I
The Licensee complies with the position in Revision 1 of this regulatory guide as presented in Subsections 11.2.3.3 and 11.3.3.7.
B/B-UFSAR A1.110-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.110 COST BENEFIT ANALYSIS FOR RADWASTE SYSTEMS FOR LIGHT-WATER-COOLED NUC LEAR POWER REACTORS The Licensee complies wi th the Annex to 10 C FR 50 Appendix I; therefore, this guide is not applicable.
B/B-UFSAR A1.111-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.111 METHODS FOR ESTIMATING ATMOS PHERIC TRANSPORT AND DISPERSION OF GASEOU S EFFLUENTS IN ROU TINE RELEASES FROM LIGHT-WATER-COOLED REACTORS
The Licensee complies wi th the requirements in R evision 1 of this guide. This is disc ussed further in Section 2.3.
B/B-UFSAR A1.112-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.112 CALCULATION OF RELEASES OF RADIOACTIVE MATERIALS IN GASEOUS AND L IQUID EFFLUENTS FROM LIGHT-WATER-COOLED POWER REACTORS
The Licensee complies with the requireme nts in Revision 0-R of this guide. This is discussed further in Se ctions 11.2 and 11.3.
B/B-UFSAR A1.113-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.113 ESTIMATING AQUATIC D ISPERSION OF EFFLUENTS FROM ACCIDENTAL AND RO UTINE REACTOR RELEASES FOR THE PURPOSE OF IMP LEMENTING APPENDIX I
The Licensee complies with the position in Revision 1 of this regulatory guide as presented in Subsections 2.4
.12, 2.4.13.3, and 11.2.3.
B/B-UFSAR A1.114-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.114 GUIDANCE TO OPERATORS AT THE C ONTROLS AND TO SENIOR OPERATORS IN THE CONTROL ROOM OF A NUCLEAR POWER UNIT The Licensee complies wi th the requirements in R evision 2 of this guide. Refer to Subsection 13.1
.2.2 for further information.
B/B-UFSAR A1.115-1 REVISION 16 - DECEMBER 2016 REGULATORY GUIDE 1.115 PROTECTION AGAINST LOW TRAJECTORY TURBINE MISSILES
The Licensee meets the objectives set forth in Revision 1 of this regulatory guide as presented in Section 3.5 and Section 10.2.3.
B/B-UFSAR A1.116-1 REVISION 15 - DECEMBER 2014 REGULATORY GUIDE 1.116 QUALITY ASSURANCE REQU IREMENTS FOR INSTA LLATION, INSPECTION, AND TESTING OF MECHANICAL EQUIPMENT AND SYSTEMS Regulatory Guide 1.116 endorsed ANSI N45.2.8 , Quality Assurance Requirements for Installation, Inspection, a nd Testing of Mechanical Equipment and Systems. NQA-1-199 4, Subpart 2.8, Quality Assurance Requirements for Installat ion, Inspection, and Testing of Mechanica l Equipment and Syst ems for Nuclear Power Plants supersedes this commitment to Reg ulatory Guide 1.116 and ANSI N45.2.8 as docu mented in the Qual ity Assurance Topical Report (NO-AA-10). Refer to Topical Report NO-AA-10 for further information on the Quali ty Assurance Program.
B/B-UFSAR A1.117-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.117 TORNADO DESIGN CLASSIFICATION
This guide is applicable only to construction permit applications docketed after May 30, 1978. The Byron/
Braidwood construction permit application w as docketed prior to this date. For a discussion of the Byron/Braidwood design, refer to Section 3.2.
B/B-UFSAR A1.118-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.118 PERIODIC TESTING OF ELECTRIC POW ER AND PROTECTION SYSTEMS
The Licensee complies wi th the regulatory po sitions in Revision 3 of this regulatory guide wit h the following exception:
Regulatory Position C Exception is taken to includ ing the signal c onditioning and actuation logic during the conduct of pe riodic RTS a nd ESFAS response time test. Implementat ion of WCAP-1403 6-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests", October 1998, al lows the allocation of bounding response time values for these portions of the protec tion channels based on engineering data.
Exception is taken to including the pressure a nd differential pressure sensors during the cond uct of periodic RTS and ESFAS response time tests.
Implementation of WCAP
-13632, Revision 2, "Elimination of Pressure Sen sor Response Time Testing Requirements", August 19 95, allows the allocat ion of bounding response time values for these portions of t he protection channels based on en gineering data.
Regulatory Position C.2 Exception is taken to the li mitations placed on the use of jumpers or any other alterations, such as lifting leads, utilized to support safety system testing.
In order to accomplish certain tests, it is necessary to use jumpers or lif ted leads to simulate desired logic circuit conditions. The safe use of these alterations is ensured with deta iled procedure s, which include independent verification of the temporary circ uit alteration and subsequent restoration.
B/B-UFSAR A1.119-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.119 SURVEILLANCE PROGRAM FOR NEW FUEL ASSEMBLY DESIGNS
This regulatory guide is not applicable to Byron and Braidwood Stations. It was withdrawn on June 23, 1977 b ecause the NRC Staff believed that fu el surveillance progra ms should be plant specific and handled on a case-by-case b asis rather than in a detailed generic manner.
B/B-UFSAR A1.120-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.120 FIRE PROTECTION GUID ELINES FOR NUCLEAR POWER PLANTS
The Licensee's positions on the regulatory pos itions in Revision 1 are provided in detail in Chap ter 3.0 of the E xelon Generation Company report, "Byron/B raidwood Stations Fire Protection Report in Response to Appen dix A of BTP APC SB 9.5-1" (current amendment).
B/B-UFSAR A1.121-1 REVISION 15 - DECEMBER 2014 REGULATORY GUIDE 1.121 BASES FOR PLUGGING DEGRA DED PWR STEAM GENERATOR TUBES
The minimum acceptable wall thicknes s at which the t ube must be removed from service by plugging satisfies the requirements of Regulatory Guide 1.121, Revision 0. The Licen see complies with the regulatory position with the followi ng comments and exceptions keyed to paragraph numbers in the position:
Position C.1
The term "unacceptable d efects" is interpreted as applying to those imperfections resu lting from service i nduced mechanical or chemical degradation of the tube walls which have penetrated to a depth in excess of t he plugging limit.
Position C.2.a(2) and C.2.a(4)
Westinghouse has doc umented its opinions on Regulatory Guide 1.121 by corporate letter that t he Licensee concurs with. A major exception to this position is the marg in of 3 against tube failure for normal opera tion. Tube failure is defined as plastic deformation of a crack to the extent that th e sides of the crack open to a nonpar allel, elliptical configuration. The tubing can sustain added internal pressure beyond those values before reaching a condition of gross failure. We have interpreted this to apply as an operating limit for the p lant and consider that it introduces a conflict to the e stablished condi tions for plant operation as identified in the plant technic al specifications. A factor of 3 is quite o ften used in A SME Code Design guidelines.
These code practices apply to the design of hardware and to the analyses done on these d esigns. Conditions which occur during operation of the equipment and which may affect the equipment so that design values no longer apply, are not di rectly addressed by the initial code requirements. That is one reas on why plant Technical Specifications have been generated to establish safe limits of operation for power station equipmen
- t. The ASME code is not applicable to the operational cri teria of steam generator tubing. Our tubing de sign and tubing in the design condition has
B/B-UFSAR A1.121-2 REVISION 15 - DECEMBER 2014 margins in excess of
- 3. In summary, we satisfy the margin of 3 if it were used in a Code se nse as new equip ment design.
Moreover, we do not believe that this margin sho uld be utilized as a limiting condition for normal operation.
Position C.2.b
In cases where sufficient inspec tion data exists to establish a degradation allowance, the rate used will be an average time rate determined from the me an of the test data.
The combined effect of these requirements would be to establish a maximum permissible primary-to-secondary leak rate which may be below the threshold of detection with current methods of measurement.
Westinghouse has determined the maxim um acceptable length of a through-wall-crack based on secondary pipe break accident loadings which are typically twice the magn itude of normal operating pressure loads.
Westinghouse will use a leak rate associated with the crack size determined on the basis of accident loadings.
Where requirements for minimum wall are mark edly different for different areas of t he tube bundle, e.g., U-bend area versus straight length in W estinghouse designs, alternate plugging limits may be established to address the varying requirements in
a manner that does n ot require unnecessary plugging of tubes.
Position C.3.e(6)
Westinghouse supplied computer code names an d references rather than the actual codes.
Nuclear Energy Institu te (NEI) 97-06 "St eam Generator Program Guidelines," as approv ed by the NRC via Lice nse Amendments 150 and 179 for Byron an d License Amendments 144 and 172 for Braidwood, contains ad ditional requirements for tube integrity in accordance with Technical Specificat ion 3.4.19.
B/B-UFSAR A1.122-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.122 DEVELOPMENT OF FLOOR DESIGN RESPONSE S PECTRA FOR SEISMIC DESIGN OF FLOOR-SUPPORTED EQUIPMENT OR COMPONENTS The plant design conforms to Rev ision 1 of this regulatory guide as described in Subsection 3.7.2.
B/B-UFSAR A1.123-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.123 QUALITY ASSURANCE REQU IREMENTS FOR CONTROL OF PROCUREMENT OF ITEMS AND SERVICES FOR NUCLEAR POWER PLANTS
The requirements in Regulato ry Guide 1.123 have been incorporated in Regulatory Guide 1.28, Revision 3. Regul atory Guide 1.123 was withdrawn on June 17, 1991.
The Licensee complies with the intent of Reg ulatory Guide 1.28 Revision 3, but applies it to ANSI/ASME NQA-1-1994.
B/B-UFSAR A1.124-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.124 SERVICE LIMITS AND LOADI NG COMBINATIONS FOR CLASS 1 LINEAR-TYPE CO MPONENT SUPPORTS The design of the Byron/Braidwood NSSS components supports is in compliance with all of the applicable regulatory positions contained in Revision 1 of this regulatory g uide. See Sections 3.6, 3.7, 3.9, and 3.10 for further information.
B/B-UFSAR A1.125-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.125 PHYSICAL MODELS FOR DESIGN A ND OPERATION OF HYDRAULIC STRUCTURES AND SYSTEMS FOR NUCLE AR POWER PLANTS This regulatory guid e is applicable to c onstruction permit applications docketed after March 1977.
The Byron/Braidwood construction permit appl ication was docketed pri or to this date.
There are no safety-related hydraulic model tests used in the design.
B/B-UFSAR A1.126-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.126 AN ACCEPTABLE MO DEL AND RELATED STAT ISTICAL METHODS FOR THE ANALYSIS OF FU EL DENSIFICATION Regulatory Guide 1.126 allows the use of NRC-approved vendor models rather than the model provided in the guide. The densification model pr esented in WCAP-8218 (proprietary) was approved by the NRC. The nonpro prietary models in WCAP-8219 and WCAP-8264 are statio n-specific models ba sed on the approved model.
B/B-UFSAR A1.127-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.127 INSPECTION OF WATER-CONTROL STRUCTURES ASSOCIATED WITH NUCL EAR POWER PLANTS The Licensee meets the requirements of the r egulatory position in Revision 1 of this guide.
B/B-UFSAR A1.128-1 REVISION 7 - DECEMBER 1998 REGULATORY GUIDE 1.128 INSTALLATION DES IGN AND INSTALLATION OF LARGE LEAD STORAGE BATTERIES FOR NUCLEAR POWER PLANTS The Licensee complies wi th the requirements in R evision 1 of this guide with the exceptions and/or clarifications to the regulatory positions identified a nd justified below:
Regulatory Position C-1 In Subsection 4.1.4, "Ve ntilation," instead of the second sentence, the fol lowing should be used:
"The ventilation system shall limit hydrogen concentration to less than two percent by volume at any location within the battery area." Licensee's Position The ventilation requ irements set for th in IEEE Std.
484-1987 are adequate.
Justification of Lic ensee's Position IEEE 484-1987 requires t hat the ventilation system limit hydrogen accum ulation to less than 2% of the total volume of the battery area.
This Regulatory Position would require that hydrogen a ccumulation be limited to less than 2% at any l ocation within the battery area. T he ventilation r equirements as set forth in IEEE 484-1987 are entirely adeq uate. The "2%
at any location" require ment would be almost impossible to verify a nd might even require the installation of ducts, vanes, and/or auxiliary fans so as to ensure that every "nook and cranny" is thoroughly purged.
The battery area ventilation system is designed to maintain the hydrogen concentr ation below 2% with a "run-away" charger (i.e., a charger delivering its full-rated output into a fully-charg ed battery, thereby causing gassing of all cells). Thus, any significant hydrogen build-up in the battery area would require two failur es (a failure of the ventilation system, and a fa ilure of the charger), both of which will be annunciated in the main control room.
B/B-UFSAR A1.128-2 REVISION 7 - DECEMBER 1998 Regulatory Position C-2 In Subsection 4.2.1, " Location," Item 1, the general requirement that the battery be prot ected against fire should be supplemented with the applicable recommendati ons in Regulatory Guide 1.120, "Fire Protection Gui delines for Nuclear Power Plants."
Licensee's Position
The reference to Regulatory Guide 1.120 is inappropriate because this regulatory guide is in the "comment" stage.
Justification of Lic ensee's Position
The battery location and pro tection against fire will be described in the Fire Protecti on Report in Res ponse to Branch Technical Position APCSB 9.5.1 in lieu of Regula tory Guide 1.120.
The location and fire protection req uirements set fo rth in IEEE 484-1987 are adequate.
In reference to Regu latory Guide 1.120, Revision 1, (November 1977), Section C.6(g), Page 20, "Safety-Rela ted Battery Rooms," Licensee's comments are as follows:
(a) This paragraph seems to imply that all safety-related batteries are to be located in separately -
enclosed rooms. It is the Licensee's po sition that it should not be ne cessary that bat tery rooms be separated from other areas of the plant by barriers having a minimum fire rating of three hours. Such barriers would be necessary only if the batteries were in a separate fire protectio n zone. There is nothing wrong with a design wherein the battery is located in an open area so long as the battery is protected from mechanical damage; e.g., the battery may be located in an electrical eq uipment room but protected by an enclosing fence.
(b) The location of d-c swit chgear and inv erters in the electrical equipment room described above is a satisfactory arrangement.
Regulatory Position C-3 Items 1 through 5 of Subsection 4.2.2, "Moun ting," should be supplemented with the following:
B/B-UFSAR A1.128-3 REVISION 5 - DECEMBER 1994 "6. Restraining channel beams and tie rods shall be electrically insulat ed from the cell c ase and shall also be in conformance with I tem 2 above rega rding moisture and acid resistance." In addition, the general require ment in Item 5 to use IEEE 344-1975 should be supplemented by Regulatory Guide 1.100, "Seismic Qualification of Electr ic Equipment for Nuclear Power Plants."
Licensee's Position
Restraining channel beams and tie rods need not be electrically insulated from t he cell case.
Justification of Lic ensee's Position The expense for the addi tion of electrical insulation to the restraining channel beams and tie rods is unwa rranted. Heat from an accident that can damage lead plates and vaporize electrolyte could also melt insulati on on restraining chan nels and tie rods.
In addition, rubber or plastic for ins ulation purposes will significantly increase the combustible f uel loading in the battery area and thus add to the fire hazard.
B/B-UFSAR A1.129-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.129 MAINTENANCE, TESTING AND REPLACEMENT OF LARGE LEAD STORAGE BATTERIES FOR NUCLEAR POWER PLANTS The Licensee complies with Revision 1 of the regulatory guide as described in the Technical Speci fications for the C&D batteries at Byron Station.
The Licensee complie s in general with the intent of the requirements in regu latory position C.1 of this guide.
However, the Licensee perfor ms a modified performance discharge test, as described in IEEE 450-1995. The modified test is performed in lieu of the service test and performance discharge test required by this regulatory guide because the discharge rate of the modified performance dischar ge test envelops the load cycle of the service test.
The Licensee complies wi th Revision 1 of the regulatory guide for the C&D batteries at Braidwo od Station with the following comments and exceptions as d escribed in the Technical Specifications.
The Licensee complie s in general with the intent of the requirements in regu latory position C.1 of this guide.
However, the Licensee perfor ms a modified performance discharge test, as described in IEEE S tandard 450-1995.
The modified test is perform ed in lieu of the service test and performance dischar ge test required by this regulatory guide, because the discharge rate of the modified performance dischar ge test envelops the load cycle of the service test.
B/B-UFSAR A1.130-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.130 SERVICE LIMITS AND L OADING COMBINATI ONS FOR CLASS 1 PLATE-AND-SHELL-TYPE COMPONENT SUPPORTS The design of Byron/Br aidwood NSSS compo nent supports is in compliance with all of the applicable regulatory positions contained in Revision 1 of this regulatory guide. Refer to Sections 3.6, 3.7, 3.9, and 3.10 for fur ther information.
B/B-UFSAR A1.131-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.131 QUALIFICATION TESTS OF ELECTRIC CABLES, FIELD SPLICES AND CONNECTIONS FOR LIGHT-WATER-COOL ED NUCLEAR POWER REACTORS The Licensee complies wi th the regulatory po sition in Revision 0 with the following c omments and exceptions keyed to paragraph numbers in the position:
- 1. Regulatory Position C-1 The position status that in li eu of Section 1.3.4.2.3 of IEEE 383, "Other Design B asis Events", the following should be used: "The remainde r of the complete spectrum of design basis events (e.g., ev ents such as a ste am line brea k) shall be considered in case they repre sent different t ypes of more severe hazards to ca ble operation." Licensee Response: All safety-r elated cable is qualified for the anticipated enviro nments detailed in Section 3.11 of the B/B-UFSAR. Steamlin e breaks are addressed in Subsection 3.6.1.3 of the B/B-UFSAR.
- 2. Regulatory Position C-10 The position states that in li eu of the first sentence of Section 2.5.4.4.1 of IEEE 383, the following should be used:
The ribbon gas burne r shall be mounted h orizontally such that the flame impinges on the spec imen midway between the tray rungs and so that the burner f ace is in front of and 4 inches from the cable and a pproximately 2 feet above the bottom of the tray." Licensee Response: The ribbon gas burner was mounted so that the burner face was in front of and 3 inches from the cable, as set forth in IEEE 383, Section 2.5.4.4.1.
- 3. Regulatory Position C-11 The position states that in li eu of Section 2.5.4.4.3 of IEEE 383 the following should be used: "Flame size will normally be achieved when the propane flow is 27.8 standard ft per hour and the air flow is 139 standard ft per hour." Licensee Response: Flame si ze was achieved using the schematic arrangement and pres sures as set for th in IEEE 383, Section 2.5.4.4.3.
B/B-UFSAR A1.132-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.132 SITE INVESTIGATIONS FOR FOUNDATIONS OF NUCLE AR POWER PLANTS
I. BYRON STATION All Byron Station site investiga tions were performed prior to June 1, 1978, with the exception of structure specific exploration consisting of eleven borings; (IC-1 through IC-11) performed on June 12, 13, and 14, 1978, t hree borings on December 14, 15, and 16, 1981, and four borings on March 17, 18, and 19, 1982, along a portion of the essential service water pipeli ne. The site in vestigations of these borings conform to the guidelines set fo rth in Revision 1 of the regulatory guide.
The site invest igations performed by the Licensee prior to the date of the regulatory guide implementation conform to the guidelines set forth in this regulatory guide because the sampling and exploration methods conform to the ASTM (American Society for Testing and Materials) procedures or other generally acc epted procedures for foundation investi gations at the time the work was performed. For details see Byron Sections 2
.4 and 2.5.
II. BRAIDWOOD STATION All Braidwood Station site investigation work was performed prior to June 1, 1978. However, the site investigations performed by the Licensee prior to the date of the regulatory guide implementation c onform to the guid elines set forth in Revision 1 of this regulatory guide in that the investigation methods conform to the ASTM (America n Society for Testing and Materials) procedures or other generally acc epted procedures for foundation investi gations at the time the work was performed. For details see Braidwood Sectio ns 2.4 and 2.5.
B/B-UFSAR A1.133-1 REVISION 4 - DECEMBER 1992 REGULATORY GUIDE 1.133 LOOSE-PART DETECTION PROGRAM FOR THE PRIMARY SYSTEM OF LIGHT-WATER-C OOLED REACTORS The loose parts detection syst em is in compliance with the regulatory position in R evision 1 with the f ollowing exceptions and clarifications keyed to paragraph numbers in the regulatory position.
Section C.l.a.
Sensor Location
Byron/Braidwood is in compliance with this section.
Section C.l.b S ystem Sensitivity
The manufacturer states that preliminary tes ts on the system demonstrate compliance with this section.
Section C.l.c. Channel Separation The sensors, cables, and lin e drivers are all physically separated from each othe r, but the twi sted-shielded pairs running back to the control ro om are routed in the same cable division.
Section C.l.d. Data Acquisition System
Byron/Braidwood is in compliance with this section.
Section C.l.e.
Alert Level
Byron/Braidwood is in compliance with this section.
Section C.l.f. Capabili ty for Sensor Channel Operability Tests There is no specific channel test for the system, but sensors on the reactor can be c hecked when the ro ds are moved.
Section C.1.g. Operability for Seismic and Environmental Conditions The loose parts monitori ng system has not be en demonstrated to be capable of performing its function following all seismic events that do not require pl ant shutdown, up to and including the operating basis eart hquake (OBE).
B/B-UFSAR A1.133-1 REVISION 10
- DECEMBER 2004 Section C.1.h. Quality of System Components The components have not been demonstrated to have a 40-year design life, but the regulatory guide permits setting up a replacement schedule to replace these it ems. The Byron/
Braidwood design permits this.
Section C.1.i. System Repair
Byron/Braidwood is in compliance with this section.
Section C.5 Technical Specification for the Loose-Part Detection System The requirements of the Loose-Pa rt Detection System were relocated from Technical Speci fications to t he Technical Requirements Manual in Braidwood Technic al Specification Amendment No. 98 and Byron Techn ical Specification Amendment No.
106 because the Loose-Part Detection System do es not meet any of the criteria for inclusi on described in 10 C FR 50.36(c)(2)(ii).
The requirement to pre pare and submit a spec ial report to the Commission if an inoperable chan nel cannot be re stored within 30 days has been revised to require the special rep ort be submitted to the Plant Operati ons Review Committee.
Section C.6 Notification of a Loose Part The requirement to not ify the Commission if the presence of a loose part is confirmed is no longer a pplicable since the requirements of the Loos e-Part Detection System were removed from Technical Specifications.
B/B-UFSAR A1.134-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.134 MEDICAL EVALUATION OF LICENSED PERSONNEL FOR NUC LEAR POWER PLANTS
The Licensee complies wi th the regulatory po sition in Revision 2 of this guide.
B/B-UFSAR A1.135-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.135 NORMAL WATER LEVEL AND DISCHARGE AT NU CLEAR POWER PLANTS
The plant design conforms to t he regulatory positions in Revision 0 of this regulatory guide as described in S ubsections 2.4.3, 2.4.8, and 2.4.11.
B/B-UFSAR A1.136-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.136 MATERIAL FOR CONCRETE CONTAINMENTS
The plant design conforms to t he regulatory position in Revision 0 as described in Appe ndix B. Regulatory Position C.2 is not applicable since grouted tendon systems are not used. Revision 0 was the current revisi on of the regulatory guide when the construction per mit was issued.
B/B-UFSAR A1.137-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.137 FUEL-OIL SYSTEMS FOR STA NDBY DIESEL GENERATORS
The Licensee complies wi th the requirements in R evision 1 of this regulatory guide with the following exceptions:
- a. For Byron, certain compo nents or parts of the Diesel Oil System which are not avail able as ASME Section III items are classified as safety-related, non-ASME.
This exception is taken because some original equipment vendors for fu el oil components are no longer ASME Section III suppli ers. All safety-related components maintain seismic qualificat ion and conform to Regulatory Guide 1.26 and 10 CFR Ap pendix B. All safety-related Diese l Oil system pip ing remains ASME Section III piping.
- b. For Braidwood, the guidance co ntained in Generic Letter 89-09 is util ized to procure it ems, originally constructed to ASME Section II I, which are no longer available as ASME Section III. The original ASME Class 3 classificati on for Diesel Oil System components and parts is maintained f or those items originally constructed to this code class.
- c. Regulatory Posit ion C.1.e(1) and (2)
Byron and Braidwood Stations perform pressure testing for those portions of the fu el oil system originally designed to Section III, Subse ction ND of the Code in accordance with the applicable Edition a nd Addenda of Section XI (including code cases) as specified in 10 CFR 50.55a(g) and the Statio n ISI Program Plan.
- d. Regulatory Position C.2 Byron and Braidwood Stat ion Technical Specification 5.5.13, Diesel Fuel Oil Testing Progra m, and Technical Requirements Manual Appe ndix M, Dies el Fuel Oil Testing Program establish and implement the requirements discussed in this regulatory position related to sampling and testing of new and stored fuel oil. Refer to Technical Specifica tion Section 5.5.13 and LCOs 3.8.1 and 3.8.3 and Technica l Requirements Manual Appendix M.
B/B-UFSAR A1.138-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.138 LABORATORY INVESTIGATIONS OF SOILS FOR ENGINEERING ANALYSIS AND DESIGN OF N UCLEAR POWER PLANTS I. BYRON STATION
All of the Byron Stati on laboratory tests on soils and rocks for determining soil and rock properties were performed prior to the regulatory guide implementation date of December 1, 1978, with the e xception of laboratory tests of soil along a portion of the ESW p ipeline in 1982. The laboratory tests performed in 1982 conform to the requirements of Regulatory Guide 1.138, Revision 0. The Licensee's laboratory investigations of soils and rocks prior to December 1, 1978 conform to the g uidelines set forth in the regulatory guide in that the ASTM (Amer ican Society for T esting and M aterials) procedures or other generally accepted p rocedures were used in performing the laboratory tes ting at the time the work was performed. For detail s, see Byron Sub sections 2.5.4 and 2.5.5. II. BRAIDWOOD STATION The only laboratory tests performed since December 1, 1978 at the Braidwood Station were on soils along a portion of the essential service water pipeline and several areas of the exterior dike embankment.
The essential service water pipeline is Safety Category I, the exterior dike embankment is non-safety-related.
The laboratory testing of so ils by the L icensee since December 1, 1978 conforms to the requirements of Regulatory Guide 1.138, Revision 0. (No rock has been tested since December 1, 1978.) The l aboratory testing of soil and rock by the Licensee prior to Decemb er 1, 1978 c onforms to the guidelines set f orth in Regulato ry Guide 1.138 in that the ASTM (American Society for Testi ng and Materials) procedures or other generally a ccepted procedures for laboratory testing of soil and rock were used at the time the work was performed. For deta ils see Braidwood Su bsections 2.5.4, 2.5.5, and 2.5.6.
B/B-UFSAR A1.139-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.139 GUIDANCE FOR RESIDUAL HEAT REMOVAL
The Licensee complies wi th the requirements in R evision 0 of this guide, in that t he Byron and Braidwood S tations are designed for safe shutdown concurrent with a single failure in one of the redundant ESF divisions. Howeve r, the Licensee defines the term safe shutdown as meaning hot s tandby.(1) Refer to Subsection 5.4.7 for further information.
(1)As defined in the Tec hnical Specifi cations.
BYRON-UFSAR A1.140-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.140 DESIGN, TESTING, AND MAINTEN ANCE CRITERIA FOR NORMAL VENTILATION EXHAUST SYSTEM AIR FILTRATION AND ADSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR P OWER PLANTS
The design of the non-safety-r elated filter sy stems meet the requirements in Revision 0 of this guide, except as noted below.
Revision 0 of Regulatory Guide 1
.140 was in effect when the operating license application was docketed. T he exceptions are keyed to paragraph numbers in the regulatory position.
la and lb - The equipment and co mponents (exclud ing charcoal, filter pads and separator pads) are designed to withstand a maximum of 40-ye ar integrated radiation dose and worst-case anticipa ted continuous service, rather than 40 years of continuo us service.
2a - All of the exhaust filter systems contain prefilters, HEPA filters, fan and associated instrumentation.
Charcoal adsorbers are o nly used when iodine is anticipated to be present, with heaters for over 70%
relative humidity air streams.
2b - The purge system exhaust filter train is designed for 43,900 cfm. Filter efficiency was tes ted at this capacity. The filter tr ain consists of two banks with a grating betwe en the filter ba nks. Each filter bank is three filters hi gh and seven wide.
2f and 3f -
Filter Housings All non-ESF filter h ousings, exclusive of the TSC and post-LOCA purge units are at negative pressure with respect to their surrounding s, and are located in
areas which are low airborne radiation environments.
Any in-leakage will not advers ely affect Appendix I releases, hence, the housing s were not l eak tested to the ANSI-N509 requir ements. However, all of the filter mounting fram es were leak tested in accordance with ANSI N510-80. The TSC unit housing is located in an area where the air borne radiation level of the room air may exceed th at of the air within the housing; however, it is
BYRON-UFSAR A1.140-2 at positive pressure with re spect to the surroundings, hence, it was not tested to ANSI-N509 re quirements.
The filter mounting fram es were leak tested in accordance with ANSI N510-80. The post-LOCA purge unit housing was leak tested to ANSI-N509-80 requirements.
2f and 3f - Ductwork All of the ductwork upstream of the non-ESF filter units is under negative pressure with respect to its surroundings. Ductw ork upstream of the filter units, except ductwo rk upstream of t he TSC filter unit, is located in areas of low airborne radioactivity. Any in-leaka ge will not adversely affect Appendix I re leases, hence, it was not tested to the ANSI-N509 requirements.
The ductwork upstream of the TSC filter unit is located in the H VAC equipment room. The quality of the equipment room e nvironment is th e same as that of the outside air which is within t he duct. Any leakage will be filtered prior to its release to the TSC environment, hence, this ductwork was not tested to the ANSI-N 509 requirements.
The following ductwo rk was leak tested to the ANSI-N509-80 requirements:
- 1. Positive pressure duct work for the laboratory exhaust filter unit outside the laboratory HVAC equipment room.
- 2. The positive pressure ductwork from the radwaste building exhaust filter unit in the auxiliary building.
- 3. The positive pressur e ductwork f rom the volume reduction system area ventilation exhaust filter that is located in the radwaste building.
- 4. The positive pre ssure ductwork from the post-LOCA purge filter unit.
- 5. All non-safety-related system ductwo rk that is required to operate and is under pressure within the control r oom boundary during an abnormal or accident condition.
- 6. TSC negative pressure duct sections outside the protected space where in-leakage would not normally be filtered.
BYRON-UFSAR A1.140-3 REVISION 12 - DECEMBER 2008 All remaining ductwork m eets the exception of ANSI N509 or has negligible impact on ALARA practices and therefore, was not leak tested.
Positive pressure radwaste building exhaust ductwork in the auxiliary buil ding was tested before the radwaste volume reduction system was put into operation. The fan peak pressure test was not performed.
For systems that have no isolati on devices, the fans are provided with high differential pressure trips or high/low flow trips.
3a - The components of the he aters are manu factured and assembled as per Section 5.5 of ANSI N50 9-76, similar to the requirements of heaters in safety-re lated filter systems, but the traceability of the components is not established as it would be in the case of safety-related heaters. Thus, no complete qualification program was done.
3e - Bubble tight dampers are not provided for TSC intake isolation.
5b - Airflow distribution and air-aerosol m ixing tests were not performed on the non-entry type filter u nits. Airflow distribution tests w ere performed on a ll entry-type filter trains to ensure that the airf low through any individual filter element does not exceed 120% of the element's rated capacity.
Filtration unit airflow capacity tests were performed at the system design pressure r ange corresponding to clean and dirty filter pressure loss es. Tests were performed at 1.25 times dirty filter cond itions to verify system stability only. Filter pressure losses for airflow capacity tests were simulated without filters in place.
The miscellaneous filter tan k vents system (VF)- lower flow limit criterion is no longer applied.
5c - Silicone sealant was used as a permanent sealant for HVAC ductwork.
HEPA filter bypass leakage is tested to less than 1%.
5d - Charcoal adsorber bypass lea kage is tested to less than 1%.
5c and 5d - Periodic testing for 1 00% recirculating systems located within reactor conta inments will be performed per ANSI N510-1980 Table 1.
In-place bypass leak age testing will n ot be performed on the Containment Charcoal Fi lter Unit Subsystem following the replacement of char coal or HEPA filters.
6a(3) Laboratory tests will be perfo rmed per the r equirements of Table 2 of Regulatory Guide 1.140 with the exception that the temperature will be 30° C, and ASTM D3803-1989 will be used to test for the methyl iodide removal efficiency.
BRAIDWOOD-UFSAR A1.140-4 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.140 DESIGN, TESTING, AND MAINTEN ANCE CRITERIA FOR NORMAL VENTILATION EXHAUST SYSTEM AIR FILTRATION AND ADSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR P OWER PLANTS
The design of the non-safety-r elated filter sy stems meet the requirements in Revision 0 of this guide, except as noted below.
Revision 0 of Regulatory Guide 1
.140 was in effect when the operating license application was docketed. T he exceptions are keyed to paragraph numbers in the regulatory position.
la and lb - The equipment and co mponents (exclud ing charcoal, filter pads and separator pads) are designed to withstand a maximum of 40-ye ar integrated radiation dose and worst-case anticipa ted continuous service, rather than 40 years of continuo us service.
2a - All of the exhaust filter systems contain prefilters, HEPA filters, fan and associated instrumentation.
Charcoal adsorbers are o nly used when iodine is anticipated to be present, with heaters for over 70%
relative humidity air streams.
2b - The purge system exhaust filter train is designed for 43,900 cfm. Filter efficiency was tes ted at this capacity. The filter tr ain consists of two banks with a grating betwe en the filter ba nks. Each filter bank is three filters hi gh and seven wide.
2f and 3f -
Filter Housings All non-ESF filter housings, exclusive of the TSC unit are at negative pressure with respect to their surroundings, and are locate d in areas which are low airborne radiation environme nts. Any inleakage will not adversely affect Appendix I releases, hence, the housings were not leak t ested to the ANSI-N509 requirements. Howev er, all of the f ilter mounting frames were leak tested in accordance with ANSI N510-80. The TSC un it housing is loca ted in an area where the airborne radiation level of the room air may exceed that of the a ir within the housing; however, it is at positive pressure with respect to the surroundings, hence, BRAIDWOOD-UFSAR A1.140-5 it was not tested to A NSI-N509 require ments. The filter mounting frames were leak tested in accordance with ANSI N510-80.
2f and 3f - Ductwork Ductwork is designed, constr ucted, and tested in accordance with the intent of Section 5.10 of ANSI N509-976. The longitudi nal seams, however, were either seal welded or mechanical lock type (Pittsburgh lock with sealants
). Silicone sealant is used as a permanent s ealant in HVAC ductwork.
All of the ductwork upstream of the non-ESF filter units is under negative pressure with respect to
its surroundings. Ductw ork upstream of the filter units, except ductwo rk upstream of t he TSC filter unit, is located in areas of low airborne
radioactivity. Any in-leaka ge will not adversely affect Appendix I re leases, hence, it was not tested to the ANSI-N509 requirements.
The ductwork upstream of the TSC filter unit is located in the H VAC equipment room. The quality of the equipment room e nvironment is th e same as that of the outside air which is within t he duct. Any leakage will be filtered prior to its release to the TSC environment, hence, this ductwork was not tested to the ANSI-N 509 requirements.
The following ductwo rk was leak test ed to the ANSI N509-80 requirements:
- 1. Positive pressure duct work for the laboratory exhaust filter unit outside the laboratory HVAC equipment room.
- 2. The positive pre ssure ductwork from the radwaste building exhaust filter unit in the
auxiliary building.
- 3. The positive pressur e ductwork f rom the volume reduction system area ventilation exhaust filter that is located in the radwaste building.
- 4. The positive pre ssure ductwork from the post-LOCA purge filter unit.
- 5. TSC negative pressure duct sections outside the protected space where in-leakage would not normally be filtered.
BRAIDWOOD-UFSAR A1.140-6 REVISION 1 - DECEMBER 1989 All remaining ductwo rk meets the exception of ANSI N509 or has negligible i mpact on ALARA practices and therefore, was not l eak tested. Positive pressure radwaste building exhaust duc twork in the auxiliary building was t ested before the radwaste volume reduction sys tem was put into operation.
Fan peak pressure tests were not performed. For systems that have an iso lation device, t he fans are provided with high diffe rential pressure trips or high/low flow trips.
3a - The components of the he aters are manu factured and assembled as per Section 5.5 of ANSI N509-76, similar to the requireme nts of heaters in safety-related filter syste ms, but the tracea bility of the components is not establishe d as it would be in the case of safety-relat ed heaters. Thu s, no complete qualification progra m was done.
3e - Bubble tight d ampers are not provided for TSC intake isolation.
5b - Airflow distribution and air-aerosol mixing tests were not performed on the non-entry type filter units. Airflow dist ribution tests were performed on all entry-type filter tra ins to ensure that the airflow through any individu al filter el ement does not exceed 120% of the element's rated capacity.
Filtration unit airflow capacity tests were performed at the system design pressure range corresponding to clean and dirty filter pressure losses. The midpoint fi lter drop test was not performed. Tests were performed at 1.25 times dirty filter conditi ons to verify system stability only. Filter pressu re losses for ai rflow capacity tests were simulated wit hout filters in place.
The miscellaneous filter tank vents system (VF), mini-purge filter unit and the post-LOCA purge filter unit (VQ) air cap acity tests were performed to verify that maximum f low is not gre ater than 110% of design.
The lower flow limit criterion was not applied.
B/B-UFSAR A1.140-7 REVISION 12 - DECEMBER 2008 5c - Silicone sealants or other temporary patching material was not used in the non-ESF fil ter housings.
Silicone sealant is used, ho wever, as a permanent sealant for HVAC ductwork.
A sampling rate of less than 1 c fm was employed for diocrylophthalate (DOP) test ing filter s ystems larger than 1000 cfm.
HEPA filter bypass l eakage is tested to less than 1%.
5d - Charcoal adsorber bypass leakage is tested to less than 1%. 5c and 5d - Periodic testing for 100% recirc ulating systems located within reactor containments wi ll be performed per ANSI N510-1980 Table 1.
In-place bypass leak age testing will n ot be performed on the Containment Charco al Filter Unit S ubsystem following the replacement of cha rcoal or HEPA filters.
6a(2) All carbon furnished pri or to 1985 as part of the original specification for atmospheric clean-up filtration units was tes ted to the req uirements of Table 5-1 of ANSI N509-1976.
All replacement carbon or original carbon f urnished in 1985 or later will be tested to the re quirements of Ta ble 5-1 of ANSI N509-1980. With the exc eption that the laboratory test for methyl iodine penetration at 30
°C, 95% relative humidity is less than 1%.
6a(3) Laboratory tests wil l be performed per the requirements of Table 2 of Regul atory Guide 1.140 with the exception that the temperature will be 30
°C and ASTM D3803-1989 will be used to test for the methyl iodide remova l efficiency.
B/B-UFSAR A1.141-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.141 CONTAINMENT ISOLATIO N PROVISIONS FOR FLUID SYSTEMS
The Licensee complies wi th the requirements in R evision 0 of this regulatory guide, as f urther explained below:
- a. Phase A and Phase B Cond itions are dif ferent from those listed. Refer to Drawings 108D685.
- b. There are differences between various figures in Appendix B and containment isolation features for various systems. Re fer to diagrams of the various systems. Appendix C per tains to diagram legend and symbols, to which the Licens ee conforms with minor exceptions. Appendix D pertains to a valve maintenance program which th e Licensee d oes not agree to implement. The Licen see agrees with the exceptions which the guide has taken to ANSI N271-1976.
B/B-UFSAR A1.142-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.142 SAFETY-RELATED CONCRETE STRUCTURES FOR NUCLE AR POWER PLANTS (OTHER THAN REACTOR VESS ELS AND CONTAINMENTS)
The Licensee is in com pliance with Revision 0 of Regulatory Guide 1.142, which was in effe ct at the time of co nstruction, with the following clarifications:
- 1. Position C.7 requires co mpliance with ANSI Standard N45.2.5-1974, i.e., two test cylinders per 100 cubic yards of concrete tested at 28 da ys with a minimum of one test per day for each class of concrete. T he Applicant's position is to take six test cylinders per 150 cubic yards of concrete tested in pairs at 7, 28, and 91 days wi th a minimum of one test per day for each class of c oncrete. This p osition is in compliance with ACI-31 8-71 and ACI-318-77.
At Byron and Braidwo od six standard cyli nders for compressive testing were prepare d from concrete sa mples representing every 150 cubic yards of concr ete placed in Category I
structures other than the co ntainment. Thes e specimens are tested for compressi ve strength at 7, 28, and 91 days.
Concrete acceptance is b ased on the 91 day results; however, the 7- and 28-day results ar e used for m onitoring the compressive strength development ages.
Requirements in ACI-318 and ACI-301 are intend ed to cover commercial structures, in which the total number of samples is small because the total volume of conc rete used is also small.
For the large volume of concrete used in a nuclear power plant, a frequency of "every 150 cu. yd." results in a much
higher confidence le vel and reliability than the "every 100 cu. yd." in ACI-301.
The rate at which concrete was p laced varied in a range of 50 cubic yards per hour up to 240 cubic yar ds per hour. This rate was governed by the size and location of the concrete element being placed and the method of placement which was used.
B/B-UFSAR A1.142-2 REVISION 6 - DECEMBER 1996 The onsite concrete batching plant h as more production quality control and lends itself to a more con sistent product than commercial concrete produced by the rea dy mix industry.
The referenced ACI-301 a nd ACI-318 requi rements have been designed for ready mix industry conditions.
The frequencies for testing fresh concrete (slump and air content) in ACI-301 and ACI-318 are 100 cubic yards and 150 cubic yards, respectively.
For Byron and Braidwood, a frequency of every 50 cubic ya rds was used f or testing slump, air content and temperature, as in Table B of ANSI N45.2.5-1974. In ad dition, the tightened sampling frequency implemented (testing of every truck) any tim e the properties of the fresh concret e were out of the allowable limits and the positive actions available to reje ct individual trucks (Table B.1-5) and to stop pr oduction (Subsection B.1.10), further reduced the prob ability that substan dard concrete was placed. ACI 349-76, "Code Requirements f or Safety-Related Concrete Structures," establishes a compressive s trength test frequency of one for every 150 cubic yards of concrete placed for safety-related s tructures other than the containment.
Section 4.3.1 of ACI 349-76 allows an increase in the number of cubic yards represe ntative of a single test by 50 cubic yards for each 100 psi lower than a standard deviation of 600 psi. Table CC-5200-1 of the Summer 1981 Adden da of the ASME Boiler and Pressure Vessel Code, Se ction III, D iv. 2 allows a testing frequency of e very 200 cubic yar ds if the average strength of at least the latest 30 consecuti ve compressive strength test exceeds the specified strength
'c f by an amount expressed as:
)69.8/f(419.1ff
'c'c cr+= (A1.142-1)
The average compressive streng th consistently exceeded this f cr for all the concrete placed.
- 2. Position C.8 requires minimu m pressure testing of embedded piping in accordance with AC I-318-71. T he Licensee's position is that all Category I embedded piping is tested in accordance with ASME Section III and all Category II embedded piping is tested in acco rdance with ANSI B31.1.
- 3. Position C.9 has been compli ed with by the Licensee.
However, the load factor for R o used in the ACI combinations 1, 2, and 3 is different tha n the load factor for R o given B/B-UFSAR A1.142-3 REVISION 4 - DECEMBER 1992 in SRP Section 3
.8.3. The load fact or used in the UFSAR combinations is in compliance with the l oad factor required by the SRP.
Load combination equations 2b' and 3b' of SR P Section 3.8.4 have been complied with by the equation numbers 6 and 5, respectively of Table 3.8-10. Note 2 of the UFSAR table when applied to equation number 6 of the UFSAR reduced this equation to equation 2b' of the SRP with the exception of the load factor for dead l oad D. The load factor used in the UFSAR is higher than the load factor used in the SRP when the seismic load and the dead load are in the same direction.
This will result in a more con servative design.
If the dead load and seismic load are not in the same direction, the load factor for D is in compliance with p osition C.11 and ACI-349-76 Section 9.3.3.
In similar manner, using Note 2 of Table 3.8-10, equation 3b' can be reduced to equa tion number 5.
The NRC review and accep tance of Subsection 3.5.1.5 (Braidwood) evaluated the ductility ratios in accordance with SRP 3.5.3 and Regulatory Guide 1.
142, Revision 1.
B/B-UFSAR A1.143-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.143 DESIGN GUIDANCE FOR RADI OACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES AND COMP ONENTS INSTALLED IN LIGHT-WATER-COOLED NUC LEAR POWER PLANTS
The Licensee complies with the requirements in Revision 0, which was in effect when t he operating license application was docketed. Further informati on is provided in Subsection 11.2.1.11.
A review of the design of the volume red uction system shows that it conforms to Regulat ory Position 1.2 with the following exceptions:
- a. Those exceptions listed in Topical Report No.
AECC-2-P(NP) inc luding amendments.
- b. No high level alarm has been p rovided for the contaminated oil tank (0VR04 T) because t he switches that allow the tank to be filled are mounted locally and it requires approximatel y 2 minutes to fill the tank. c. No high level alarm has been pro vided for the flush water recovery tank (0 VR09T). The h igh level switch starts the flush wat er recovery tank p ump (0VR30M).
B/B-UFSAR A1.144-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.144 AUDITING OF QUALITY AS SURANCE PROGRAMS FOR NUCLEAR POWER PLANTS The requirements in Regulato ry Guide 1.144 have been incorporated in Regulatory Guide 1.28, Revision 3. Regul atory Guide 1.144 was withdrawn on June 17, 1991.
The Licensee complies with the intent of Reg ulatory Guide 1.28 Revision 3, but applies it to ANSI/ASME NQA-1-1994.
B/B-UFSAR A1.145-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.145 ATMOSPHERIC DISPERSI ON MODELS FOR PO TENTIAL ACCIDENT CONSEQUENCE ASSESSMENTS AT NUCLEAR POWER PLANTS Regulatory Guide 1.145 h ad not been issued at the time of the original SAR submitt al. Values for /Q were calcula ted using the NRC guidance available.
This guidance was provided in Section 2.3.4.2 of the S tandard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Revision 1, issued October 1972. Values for /Q were calculated using the observed onsite meteorological record, and the 5% and 50% pr obability values were selected and reported in the SAR.
This approach was the common practice at the time, and it was reviewed and ac cepted by the NRC. The NRC published Revisi on 0 of Regulatory G uide 1.145 in August 1979. The methodolo gy for calculating /Q differs from the older methods, but results in values of /Q similar to those calculated earlier. The NRC does not requi re licensees u sing the older method to recalculate the /Q values using Regulatory Guide 1.145.
B/B-UFSAR A1.146-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.146 QUALIFICATION OF QUALITY ASSURANCE PROGRAM AUDIT PERSONNEL FOR NU CLEAR POWER PLANTS The requirements in Regulato ry Guide 1.146 have been incorporated in Regulatory Guide 1.28, Revision 3. Regul atory Guide 1.146 was withdrawn on June 17, 1991.
The Licensee complies with the intent of Reg ulatory Guide 1.28 Revision 3, but applies it to ANSI/ASME NQA-1-1994.
B/B-UFSAR A1.147-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.147 INSERVICE INSPECTION C ODE CASE ACCEPTABILITY ASME SECTION XI, DIVISION 1 This regulatory guid e lists those Section XI ASME code cases that are generally acceptab le to the NRC for impl ementation in the inservice inspection of light-water-cooled nuc lear power plants.
The Licensee complie s with the regulat ory position.
Code cases expla in the intent of cod e rules or provide alternative requirements und er special c onditions.
Implementation of indi vidual code cases is limited to the requirements as specified in the inquiry and reply sections of each code case. The ASME considers the use of code cases to be optional for the user and not a mandatory requirement. Use of this regulatory guide is optional.
Approval of code cases l isted in this regulatory guide is by code case number and date of ASME approval. Code cases to be applied during an inspection interval or preservice inspection that were previously approved for use need not be changed because a subsequent revision of the code case is listed as the approved version in subsequent revisions of this regu latory guide.
Similarly, code cases to be applied during an inspection interval or preservice inspection that we re previously ap proved for use need not be changed beca use the code case ha s been subsequently annulled. A code case that was approved for a particular situation and not for a generic application should be used only for the approved situati on, because annulment of such a code case could result in situatio ns that would not meet code requirements.
New revisions to code cases must be accepted by the NRC prior to their use.
Code cases listed in this re gulatory guide are generically acceptable for imple mentation in the i nservice inspection program. Beginning with Revision 6 to this regu latory guide, it is no longer necessary to obtain NRC approval to use code cases listed in the regulatory guide. Use of such code cases should be noted in the applicable inservice inspection p rogram plan and/or procedures.
Code cases not l isted in this regulato ry guide cannot be implemented in the ins ervice inspection program unless formal written approval of a specific relief reques t is obtained from the Commission or other formal approval is obtained from the Commission in accordance with footnote 6 of the Codes and Standards Rule, 10CFR 50.55a.
B/B-UFSAR A1.149-1 REVISION 14 - DECEMBER 2012 REGULATORY GUIDE 1.149 NUCLEAR POWER PLANT SIMULATION FACILITIES FOR USE IN OPERATOR TRAINING , LICENSE EXAMINATIO NS, AND APPLICANT EXPERIENCE REQUIREMENTS
The requirements of ANSI-ANS-3.5-2009 and the clarifications to that document contained in Revision 4 of Reg ulatory Guide 1.149, Section C have been in corporated into the Exelon Nuclear Training procedures that cover maintenance of s imulator fidelity and configuration management. T he requirements and actions prescribed in these pr ocedures have been imple mented at the Byron and Braidwood simulators.
Revision 4 of the regulatory guide, issued April 2011, r equires that operating tests be administered on an approved or ce rtified simulator.
B/B-UFSAR A1.150-1 REVISION 13 - DECEMBER 2010 REGULATORY GUIDE 1.150 ULTRASONIC TESTING OF REACTO R VESSEL WELDS DURING PRESERVICE AND INSER VICE EXAMINATIONS On February 4, 2008, Regulatory Guide 1.
150 requirements for ultrasonic testing of reactor vessel welds w ere superseded by 10 CFR 50.55a(g)(6)(ii)(C)(1).
Ultrasonic testing of reactor vessel welds are conducted using any of the following:
- 1. Welds that are s pecified by 10 CFR 50.55a to be examined using the requirements of American Society of Mechanical Engineers (ASME) Section XI, A ppendix VIII or the Appendix VIII requirements are modifi ed by 10 CFR 50.55a or;
- 2. Welds specified by ASME Sect ion XI to be e xamined using the requirements of Append ix VIII or are demonstrated as an acceptable alternative using the methods described in ASME Section XI, IWA-2240 (as allowed by 10 CFR 50.55a) or;
- 3. Welds allowed by ASME Nuclear Code Cases to be examined using the requirements of Appendix VIII or alternate m ethods. These code cases are implemented using approved relief requests or are approved for use in Regulatory Guide 1.147 or;
- 4. Welds allowed by NRC approved relief requests to be examined using the requirements of an alternative method.
B/B-UFSAR A1.151-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.151 INSTRUMENT SENSING LINES
Regulatory Guide 1.151 is not ap plicable to Byron and Braidwood stations because the construction permit application was issued before the imple mentation date.
B/B-UFSAR A1.153-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.153 CRITERIA FOR POWER, INST RUMENTATION, AND CONTROL PORTIONS OF SAFETY SYSTEMS Regulatory Guide 1.153 is not ap plicable to Byron and Braidwood stations because the construction permit application was docketed before the imple mentation date.
B/B-UFSAR A1.154-1 REVISION 7 - DECEMBER 1998 REGULATORY GUIDE 1.154 FORMAT AND CONTENT OF PL ANT-SPECIFIC PRESSURIZED THERMAL SHOCK SA FETY ANALYSIS REPORTS FOR PRESSURIZED WATER REACTORS
The stated purpose of Regulatory Guide 1.154 is to provide
recommended methods of a ssessing the risk due to pressurized thermal shock (PTS) even ts for proposed operation of the plant with reactor vessel RTPTS above the scr eening criteri
- a. Since the reactor vessels do not exceed the NRC PTS screening criteria for both design and exte nded license vessel life, Regulatory Guide 1.154 is not applicable.
B/B-UFSAR A1.155-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.155 STATION BLACKOUT
The plant design conforms to the regulatory position described in Regulatory Guide 1.155, Revision 0. This is discussed further in Subsection 8.3.1.1.2.2.
B/B-UFSAR A1.156-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.156 ENVIRONMENTAL QUALIF ICATION OF CONNECT ION ASSEMBLIES FOR NUCLEAR POWER PLANTS Regulatory Guide 1.156 is not ap plicable to Byron and Braidwood stations due to the dates that the con struction permit was issued and the operating license ap plication was docketed.
B/B-UFSAR A1.157-1 REVISION 9 - DECEMBER 2002 REGULATORY GUIDE 1.157 BEST-ESTIMATE CALCULATIONS OF EMERGENCY CORE COOLING SYSTEM PERFORMANCE Byron Station and Br aidwood Station were subsequently licensed via License Amendment No
- s. 118 and 112, respectively, to allow use of the generically approved Westinghouse Best-Estimate large break loss-of-coolant ac cident (LBLOCA) analys is methodology as the methodology used to perform LBLOCA analy ses for Byron and Braidwood Stations. This me thodology is described in WCAP-12945-P-A and was approved by the NRC in a letter from R. C.
Jones, NRC, to N. J. L iparulo, Westinghouse El ectric Corporation, "Acceptance for Referenc ing of the Topical R eport WCAP-12945(P), 'Westinghouse Code Qua lification Document Fo r Best Estimate Loss of Coolant Analysis,'"
dated June 28, 1996.
Therefore, the Byron Station and Braidwood Station specific analy sis satisfies the acceptance criteria of 10 CFR 50.46, conforms to the requirements of 10 CFR 50, Appendix K, "ECCS Evaluation Mod els," Section II, "Required Documentation," and meets the guidan ce of Regulatory Guide 1.157, "Best-E stimate Calculations of Emergency Core Cooling System Performan ce," dated May 1989.
The method of analysis for large break is di scussed in Section 15.6.5.2.1.2.
B/B-UFSAR A1.158-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 1.158 QUALIFICATION OF SAFET Y-RELATED LEAD STO RAGE BATTERIES FOR NUCLEAR POWER PLANTS Replacement safety-rel ated lead storage batteries purchased subsequent to February 28, 1989 are qualifie d in accordance with the provisions of IEEE Standard 535-1986, wh ich is endorsed by Regulatory Guide 1.1 58, Revision 0.
B/B-UFSAR A1.159-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 1.159 ASSURING THE AVAILABILITY OF FUNDS FOR DECOMMISSIONING NUCLEAR REACTORS Decommissioning costs include the cost of decontamination, dismantling, and site restoration in accordance with NRC guidelines. Illinois law requires public utility operators of nuclear power plants to establish external t rusts to hold funds to cover the costs of the eventual decommissioning of nuclear power plants. The I llinois Commerce Com mission has approved Exelon Generation Comp any's method of fundin g its obligations with respect to deco mmissioning costs and required Exelon Generation Company to contribu te future decomm issioning fund collections to the t rusts annually.
The guidance provided in Regulatory Guide 1.
159 was issued after the Licensee's plan was approved.
B/B-UFSAR A1.160-1 REVISION 7 - DECEMBER 1998 REGULATORY GUIDE 1.160 MONITORING THE EFFECTIVE NESS OF MAINTENANCE AT NUCLEAR POWER PLANTS The Licensee complies with the requirements in Regulatory Guide 1.160, Revision 2, through the implement ation of 10 CFR 50.65.
The regulatory guide endorses NUMARC 93-01, "Industry Guideline for Monitoring the Eff ectiveness of Maintenance at Nuclear Power Plants," Revision 2, as an accep table method for implementing the maintenance rule.
The maintenance rule h as been implemented using the referenced documents, with the fo llowing exceptions and clarifications:
NUMARC 93-01 rec ommends that ind ustry-wide opera ting experience be reviewed for plant-specific applicability during the scoping process. Events that have o ccurred at simil arly configured plants should be consi dered to identify non-safety related systems, structures, and compone nts (SSC) that m eet the scoping criteria. Specific ev ents were not directly reviewed during the scoping process. Indu stry operating exp erience was considered indirectly in the scoping process via the expert panelists with operating licenses a nd system engineering experience. These panelists programmatically rec eive all perti nent operating experiences through routine routing and requalification training.
The expert panel used their know ledge of this indust ry experience to answer the scopin g screening criteria.
In addition, where necessary, the e xpert panel considered r easonable hypothetical scenarios to determine if an SSC met the scoping criteria, even if a specific event was not identified to veri fy inclusion in the scope of the rule.
B/B-UFSAR A1.160-2 REVISION 7 - DECEMBER 1998 NUMARC 93-01 recomme nds monitoring mai ntenance preventable functional failures (MPFFs). Byron St ation monitors all functional failures and does not classify certain failures as "maintenance preventable."
The classificati on "maintenance preventable" is often su bjective and could res ult in inconsistent evaluations for the need for (a)(1) goal setting.
The use of all functional failures is more conservative than the recommendations of NUMARC 93-01 (use of MPFFs) in that all functional failures, not just those attributed to m aintenance related reasons, are considered when eval uating the need for goals. Though conservative, this is considered to result in mo re effective implementation because it focuses on fix ing performance problems rather than categorizing them. This approac h is also consistent with the unavailability monitoring, which tr acks unavailability due to all cause s, not just those associ ated with maintenance activities. In additi on, monitoring all functional failures simplifies the m onitoring process. It removes the subjective aspect of MPFF determina tion, which often prov ides little value in selecting appropriate cor rective actions. It is also consistent with the ne eds of the probabi listic risk assessment group, which uses pl ant-specific failure dat a attributed to all causes, not just MPFFs. During the pilot inspections, the NRC approved of this approach.
NUMARC 93-01 rec ommends that "The hist orical data used to determine the performance of SSCs consists of that data for a period of at least t wo fuel cycles or 36 months, whichever is less." In several cas es involving the initial evaluation of SSC performance, the historical data sou rces for SSC availability and reliability were not amenable to the exact assessment of performance. Information needed to make a corre ct maintenance rule determination may not have been documen ted. Consequently, historical information dating from t he start of cycle 6 for Unit 1 and cycle 5 for Un it 2 for Byron was used to the extent possible. Candidates for the (a
)(1) category were based on this review and the r ecommendations by the site maint enance rule owner and senior stati on management.
The details of maintenan ce rule implementation and compliance are described in station procedures.
B/B-UFSAR A1.161-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.161 EVALUATION OF REACTOR PR ESSURE VESSELS WITH CHARPY UPPER-SHELF ENERGY L ESS THAN 50 FT-LB Regulatory Guide 1.1 61 does not apply si nce the Charpy upper-shelf energy is predicted to remain above 50 ft-lb.
Appendix G to 10 CFR 50 requires that the predicted Charpy upper-shelf energy at end of life be above 50 ft-l
- b. Using the method in Regulatory Guide 1.99, Revision 2, the pr edicted Charpy upper-shelf energy of the weld met al at the end of l ife will be greater than 50 ft-lb.
B/B-UFSAR A1.162-1 REVISION 6 - DECEMBER 1996 REGULATORY GUIDE 1.162 FORMAT AND CONTENT OF RE PORT FOR THERMAL ANNEALING OF REACTOR PRE SSURE VESSELS Regulatory Guide 1.162 d oes not apply since there is no intention to perform thermal a nnealing of the reac tor pressure vessels.
B/B-UFSAR A1.163-1 REVISION 12 - DECEMBER 2008 REGULATORY GUIDE 1.163 PERFORMANCE-BASED CONTAI NMENT LEAK-TEST PROGRAM The Licensee follows the guidance in Revision 0 of Regulatory Guide 1.163 as modified by approved exce ptions in Technical Specification 5.5.16 for a per formance-based l eak-test program and leakage-rate test me thods, procedures an d analyses that are used to comply with the performance-based Op tion B in Appendix J to 10 CFR Part 50.
B/B-UFSAR A1.165-1 REVISION 7 - DECEMBER 1998 REGULATORY GUIDE 1.165 IDENTIFICATION AND C HARACTERIZATION OF S EISMIC SOURCES AND DETERMINATION OF SAFE SHUTDOWN EARTHQUAKE GROUND MOTION This guide is applicable only to applications for construction permits, operating licenses, combined licenses, or design certifications submitt ed after January 1 0, 1997. The Byron/Braidwood documents were submitted pri or to this date.
B/B-UFSAR A1.166-1 REVISION 7 - DECEMBER 1998 REGULATORY GUIDE 1.166 PREEARTHQUAKE PLANNING A ND IMMEDIATE NUC LEAR POWER PLANT OPERATOR POSTEAR THQUAKE ACTIONS This guide is applicable only to applications for construction permits, operating licenses, combined licenses, or design certifications submitt ed after January 1 0, 1997. The Byron/Braidwood documents were submitted pri or to this date.
B/B-UFSAR A1.167-1 REVISION 7 - DECEMBER 1998 REGULATORY GUIDE 1.167 RESTART OF A NUCLEAR POWER PLANT SHUT DOWN BY A SEISMIC EVENT This guide is applicable only to applications for construction permits, operating licenses, combined licenses, or design certifications submitt ed after January 1 0, 1997. The Byron/Braidwood documents were submitted pri or to this date.
B/B-UFSAR A1.181-1 REVISION 10 - DECEMBER 2004 REGULATORY GUIDE 1.181 CONTENT OF THE UPDAT ED FINAL SAFETY AN ALYSIS REPORT IN ACCORDANCE WITH 10 CFR 50.71(e)
As part of the ongoing effort to improve the quality of the UFSAR, the guidelines pr ovided in Nuclear En ergy Institute (NEI) 98-03, "Guidelines for Updating Fina l Safety Ana lysis Reports,"
Revision 1, June 1999, as endorsed by NRC Regulatory Guide 1.181, "Content of the Updated Final Safety Ana lysis Report in Accordance with 10CFR5 0.71(e)," Revision 0, September 1999, are used to further improve the content of the UFSAR. While the UFSAR will continue to follow the gene ral organizational recommendations, i.e., format, specified in Regulatory Guide 1.70, Revision 2, the reorganization options described in NEI 98-03 will be used to simplify informati on contained in the UFSAR to improve its f ocus, clarity, and m aintainability.
B/B-UFSAR A1.183-1 REVISION 12 - DECEMBER 2008 REGULATORY GUIDE 1.183 ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR EVAL UATING DESIGN BASIS ACCIDENTS AT NUCLEAR POWER REACTORS
The Licensee complies with Revision 0 of the regulatory position with comments and ex ceptions as listed in the following UFSAR Tables:
- Table A1.183-1
- Table A1.183-2
- Table A1.183-3
- Table A1.183-4
- Table A1.183-5
- Table A1.183-6
- Table A1.183-7
B/B-UFSAR A1.183-2 REVISION 12 - DECEMBER 2008 TABLE A.1.183-1 CONFORMANCE WITH REGULATORY GUIDE 1.183 MAIN SECTIONS RG Section RG Position AnalysisComments3.1 The inventory of fission products in the reactor core and available for release to the containment should be based on the maximum full power operation of the core with, as a minimum, current licensed values for fuel enrichment, fuel burnup, and an assumed core power equal to the current licensed rated thermal power times the ECCS evaluation uncertainty. The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. The core inventory should be determined using an appropriate isotope generation and depletion computer code such as ORIGEN 2 or ORIGEN-ARP. Core inventory factors (Ci/MWt) provided in TID 14844 and used in some analysis computer codes were derived for low burnup, low enrichment fuel and should not be used with higher burnup and higher enrichment fuels.
ConformsORIGEN 2.1 based methodology was used to determine the bounding core inventory. These source terms were evaluated at end-of-cycle and at beginning of cycle (100 effective full power days (EFPD) to achieve equilibrium) conditions. The worst-case inventory was used for each of the selected 60 isotopes for the RADTRAD analyses. These values were then converted to units of Ci/MWt. Accident analyses are based on a 3658.3 MWt power level, based on the current accident analysis design basis allowance for instrument uncertainty. Source terms are based on an 18-month fuel cycle with 542.9 EFPD per cycle.3.1 For the DBA LOCA, all fuel assemblies in the core are assumed to be affected and the core average inventory should be used. For DBA events that do not involve the entire core, the fission product inventory of each of the damaged fuel rods is determined by dividing the total core inventory by the number of fuel rods in the core. To account for differences in power level across the core, radial peaking factors from the facility's core operating limits report (COLR) or Technical Specifications should be applied in determining the inventory of the da maged rods. ConformsPeaking factors of 1.7 are used for DBA events that do not involve the entire core, with fission product inventories for damaged fuel rods determined by dividing the total core inventory by the number of fuel rods in the core. 3.1 No adjustment to the fission product inventory should be made for events postulated to occur during power operations at less than full rated power or those postulated to occur at the beginning of core life. For events postulated to occur while the facility is shutdown, e.g., a fuel handling accident, radioactive decay from the time of shutdown may be modeled. ConformsNo adjustments for less than full power are made in any analyses.
B/B-UFSAR A1.183-3 REVISION 12 - DECEMBER 2008 TABLE A.1.183-1 CONFORMANCE WITH REGULATORY GUIDE 1.183 MAIN SECTIONS RG Section RG Position AnalysisComments3.2 The core inventory release fractions, by radionuclide groups, for the gap release and early in-vessel damage phases for DBA LOCAs are listed in Table 1 for BWRs and Table 2 for PWRs. These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1. TABLE 2 PWR Core Inventory Fraction Released Into Containment Gap Early Release In-Vessel Group Phase Phase Total Noble Gases 0.05 0.95 1.0 Halogens 0.05 0.35 0.4 Alkali Metals 0.05 0.25 0.3 Tellurium Metals 0.00 0.05 0.05 Ba, Sr 0.00 0.02 0.02 Noble Metals 0.00 0.0025 0.0025 Cerium Group 0.00 0.0005 0.0005 Lanthanides 0.00 0.0002 0.0002 Footnote 10: The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak rod burnup up to 62,000 MWD/MTU. The data in this section may not be applicable to cores containing mixed oxide (MOX) fuel. ConformsThe release fractions from Regulatory Position 3.1, Table 2 are used.
Footnote 10 criteria are met. 3.2 For non-LOCA events, the fractions of the core inventory assumed to be in the gap for the various radionuclides are given in Table 3. The release fractions from Table 3 are used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor.
Table 3 11 Non-LOCA Fraction of Fission Product Inventory in Gap Group Fraction I-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens 0.05 Alkali Metals 0.12Exception taken (as approved in a previous submittal by another Licensee) The analysis does not fully comply with Note 11 of Table 3 since typical Byron and Braidwood core designs indicate that there are fuel assemblies that exceed the 6.3 kW/ft while >54GWD/MTU. Previous analyses (ANS 5.4) for TMI-1 have shown that those fuel assemblies exceeding these limits had no increase in gap release fractions of concern. Therefore, doubling of the B/B-UFSAR A1.183-4 REVISION 12 - DECEMBER 2008 TABLE A.1.183-1 CONFORMANCE WITH REGULATORY GUIDE 1.183 MAIN SECTIONS RG Section RG Position AnalysisComments Footnote 11: The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for rods with burnups that exceed 54 GWD/MTU. As an alternative, fission gas release calculations performed using NRC-approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load. For the BWR rod drop accident and the PWR rod ejection accident, the gap fractions are assumed to be 10% for iodines and noble gases. gap fractions in Table 3 is conservative as used and approved in the Fort Calhoun AST submittal.
Peaking factor of 1.7 used for DBA events that do not involve the entire core. 3.3 Table 4 tabulates the onset and duration of eachsequential release phase for DBA LOCAs at PWRs and BWRs. The specified onset is the time following the initiation of the accident (i.e., time = 0). The early in-vessel phase immediately follows the gap release phase. The activity released from the core during each release phase should be modeled as increasing in a linear fashion over the duration of the phase. For non-LOCA DBAs, in which fuel damage is projected, the release from the fuel gap and the fuel pellet should be assumed to occur instantaneously with the onset of the projected damage. Table 4 LOCA Release Phases PWRs BWRs Phase Onset Duration Onset Duration Gap Release 30 sec 0.5 hr 2 min 0.5 hr Early In-Vessel 0.5 hr 1.3 hr 0.5 hr 1.5 hr ConformsThe PWR durations from Table 4 are used.
The LOCA activity released from the core is modeled in a linear fashion over the duration of the release phases.
Non-LOCA DBAs are modeled as an instantaneous release from the fuel. 3.3 For facilities licensed with leak-before-break methodology, the onset of the gap release phase may be assumed to be 10 minutes. A licensee may propose an alternative time for the onset of the gap release phase, based on facility-specific calculations using suitable analysis codes or on an accepted topical report shown to be applicable for the Not ApplicableNeither Byron nor Braidwood use leak-before-break methodology for design bases dose analyses.
B/B-UFSAR A1.183-5 REVISION 12 - DECEMBER 2008 TABLE A.1.183-1 CONFORMANCE WITH REGULATORY GUIDE 1.183 MAIN SECTIONS RG Section RG Position AnalysisComments specific facility. In the absence of approved alternatives, the gap release phase onsets in Table 4 should beused.3.4 Table 5 lists the elements in each radionuclide group that should be considered in design basis analyses. Table 5 Radionuclide Groups Group Elements Noble Gases Xe, Kr Halogens I, Br Alkali Metals Cs, Rb Tellurium Group Te, Sb, Se, Ba, Sr Noble Metals Ru, Rh, Pd, Mo, Tc, Co Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am Cerium Ce, Pu, Np ConformsThe nuclides used are the 60 identified as being potentially important dose contributors to total effective dose equivalent (TEDE) in the RADTRAD code, which encompasses those listed in RG 1.183, Table 5. The Co-58 and Co-60 values used are those from the RADTRAD defaults (activation products). All other isotope activities were determined using ORIGEN.3.5 Of the radioiodine released from the reactor coolant system (RCS) to the containment in a postulated accident, 95 percent of the iodine released should be assumed to be cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. This includes releases from the gap and the fuel pellets. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form. The same chemical form is assumed in releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs other than FHAs or LOCAs. However, the transport of these iodine species following release from the fuel may affect these assumed fractions. The accident-specific appendices to this regulatory guide provide additional details.ConformsThis guidance was applied in the analyses.
(95 percent of the iodine released should be assumed to be cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form.)3.6 The amount of fuel damage caused by non-LOCA design basis events should be analyzed to determine, for the case resulting in the highest radioactivity release, the fraction of the fuel that reaches or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for which the fuel clad is breached. Although the NRC staff has traditionally relied upon the departure from nucleate boiling ratio (DNBR) as a fuel damage criterion, licensees may propose other methods to the NRC staff, such as those based upon enthalpy deposition, for estimating fuel damage ConformsThe currently licensed and approved assumptions regarding the amount of fuel damage for non-LOCA design basis events is used in the AST analyses.
B/B-UFSAR A1.183-6 REVISION 12 - DECEMBER 2008 TABLE A.1.183-1 CONFORMANCE WITH REGULATORY GUIDE 1.183 MAIN SECTIONS RG Section RG Position AnalysisComments for the purpose of establishing radioactivity releases.4.1.1 The dose calculations should determine the TEDE. TEDE is the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE) from external exposure. The calculation of these two components of the TEDE should consider all radionuclides, including progeny from the decay of parent radionuclides that are significant with regard to dose consequences and the released radioactivity. ConformsTEDE is calculated, with significant progeny included. 4.1.2 The exposure-to-CEDE factors for inhalation of radioactive material should be derived from the data provided in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers" (Ref. 19). Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20), provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the CEDE. ConformsFederal Guidance Report 11 dose conversion factors (DCFs) are used. 4.1.3 For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite should be assumed to be 3.5 x 10
-4 cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate should be assumed to be 1.8 x 10
-4 cubic meters per second. After that and until the end of the accident, the rate should be assumed to be 2.3 x 10
-4cubic meters per second.ConformsThe values that correspond to the rounded values in Section 4.1.3 of RG 1.183 are used. 4.1.4 The DDE should be calculated assuming submergence in semi-infinite cloud assumptions with appropriate credit for attenuation by body tissue. The DDE is nominally equivalent to the effective dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE. Table III.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21), provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.ConformsFederal Guidance Report12 conversion factors are used. 4.1.5 The TEDE should be determined for the most limiting person at the EAB. The maximum EAB TEDE for any two-hour period following the start of the radioactivity release should be determined and used in determiningConformsThe maximum two-hour LOCA EAB dose starts as follows:
B/B-UFSAR A1.183-7 REVISION 12 - DECEMBER 2008 TABLE A.1.183-1 CONFORMANCE WITH REGULATORY GUIDE 1.183 MAIN SECTIONS RG Section RG Position AnalysisComments compliance with the dose criteria in 10 CFR 50.67. The maximum two-hour TEDE should be determined by calculating the postulated dose for a series of small time increments and performing a "sliding" sum over the increments for successive two-hour periods. The maximum TEDE obtained is submitted. The time increments should appropriately reflect the progression of the accident to capture the peak dose interval between the start of the event and the end of radioactivity release (see also Table 6).
Footnote 14: With regard to the EAB TEDE, the maximum two-hour value is the basis for screening and evaluation under 10 CFR 50.59. Changes to doses outside of the two-hour window are only considered in the context of their impact on the maximum two-hour EAB TEDE. Containment Leakage: 11.01 rem TEDE 0.3 to 2.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ECCS Leakage: 1.20 rem TEDE 1.8 to 3.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Conservatively, the maximum 2-hour period dose was determined by adding the maximum 2-hour dose for each of the components listed above even though they do not occur simultaneously.4.1.6 TEDE should be determined for the most limiting receptor at the outer boundary of the low population zone (LPZ) and should be used in determining compliance with the dose criteria in 10 CFR 50.67.ConformsThis guidance is applied in the analyses through the use of the RADTRAD computer code.4.1.7 No correction should be made for depletion of the effluent plume by deposition on the ground.ConformsNo such corrections are made in the analyses.4.2.1 The TEDE analysis should consider all sources of radiation that will cause exposure to control room personnel. The applicable sources will vary from facility to facility, but typically will include: Contamination of the control room atmosphere by the intake or infiltration of the radioactive material contained in the radioactive plume released from the facility, Contamination of the control room atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope, Radiation shine from the external radioactive plume released from the facility, Radiation shine from radioactive material in the reactor containment, Radiation shine from radioactive material in systems and components inside or external to the control room envelope, e.g., radioactive material buildup in recirculation filters. ConformsThe principal source of dose within the control room is due to airborne activity within the CR. The dose contributions from the other sources, such as direct shine, were also considered.
B/B-UFSAR A1.183-8 REVISION 12 - DECEMBER 2008 TABLE A.1.183-1 CONFORMANCE WITH REGULATORY GUIDE 1.183 MAIN SECTIONS RG Section RG Position AnalysisComments4.2.2 The radioactive material releases and radiation levels used in the control room dose analysis should be determined using the same source term, transport, and release assumptions used for determining the EAB and the LPZ TEDE values, unless these assumptions would result in non-conservative results for the control room. ConformsThe source term, transport, and release methodology is the same for both the control room and offsite locations. 4.2.3 The models used to transport radioactive material into and through the control room, and the shielding models used to determine radiation dose rates from external sources, should be structured to provide suitably conservative estimates of the exposure to control room personnel.ConformsThis guidance is applied in the analyses. 4.2.4 Credit for engineered safety features that mitigate airborne radioactive material within the control room may be assumed. Such features may include control room isolation or pressurization, or intake or recirculation filtration. Refer to Section 6.5.1, "ESF Atmospheric Cleanup System," of the SRP (Ref. 3) and Regulatory Guide 1.52, "Design, Testing, and Maintenance Criteria for Post-accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants" (Ref. 25), for guidance. ConformsEngineered safety features that mitigate airborne radioactive material within the control room are credited. These features are qualified and acceptable per the referenced guidance. Control Room and intake and recirculation filtration are credited. Radiation isolation mode has been analyzed with initiation within 30 minutes. After this period, credit is taken for HEPA and charcoal adsorber efficiencies.4.2.5 Credit should generally not be taken for the use of personal protective equipment or prophylactic drugs. Deviations may be considered on a case-by-case basis.ConformsSuch credits are not taken. 4.2.6 The dose receptor for these analyses is the hypothetical maximum exposed individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10-4 cubic meters per second.ConformsStandard occupancy factors and breathing rate are used throughout the analyses.
B/B-UFSAR A1.183-9 REVISION 12 - DECEMBER 2008 TABLE A.1.183-1 CONFORMANCE WITH REGULATORY GUIDE 1.183 MAIN SECTIONS RG Section RG Position AnalysisComments4.2.7 Control room doses should be calculated using dose conversion factors identified in Regulatory Position 4.1 above for use in offsite dose analyses. The DDE from photons may be corrected forthe difference between finite cloud geometry in the control room and the semi-infinite cloud assumption used in calculating the dose conversion factors. The following expression may be used to correct the semi-infinite cloud dose, DDE, to a finite cloud dose, DDEfinite , where the control room is modeled as a hemisphere that has a volume, V, in cubic feet, equivalent to that of the control room (Ref. 22).
1173338.0VDDE DDE finite= ConformsThe equation given is utilized for finite cloud correction when calculating external doses due to the airborne activity inside the control room. 4.3 The guidance provided in Regulatory Positions 4.1 and4.2 should be used, as applicable, in re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in NUREG-0737 (Ref. 2). Design envelope source terms provided in NUREG-0737 should be updated for consistency with the AST. In general, radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of TEDE. Integrated radiation exposure of plant equipment should be determined using the guidance of Appendix I of this guide. ConformsTSC habitability has been re-determined using AST and has been determined acceptable. The EOF is sufficiently far away from the site (outside the LPZ) such that analysis is not required.
5.1.1 The evaluations required by 10 CFR 50.67 are re-analyses of the design basis safety analyses and evaluations required by 10 CFR 50.34; they are considered to be a significant input to the evaluations required by 10 CFR 50.92 or 10 CFR 50.59. These analyses should be prepared, reviewed, and maintained in accordance with quality assurance programs that comply with Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50. ConformsThese analyses were prepared as specified in the guidance. These analyses have been prepared and reviewed in accordance with a quality assurance program that complies with Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part50.
B/B-UFSAR A1.183-10 REVISION 12 - DECEMBER 2008 TABLE A.1.183-1 CONFORMANCE WITH REGULATORY GUIDE 1.183 MAIN SECTIONS RG Section RG Position AnalysisComments5.1.2 Credit may be taken for accident mitigation features that are classified as safety-related, are required to be operable by Technical Specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed. Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences. Conforms Accident mitigation features credited in these analyses are classified as safety-related, are required to be operable by Technical Specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. Single active failures and loss of offsite power were also considered where required.5.1.3 The numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67 should be selected with the objective of determining a conservative postulated dose. In some instances, a particular parameter may be conservative in one portion of an analysis but be non-conservative in another portion of the same analysis. ConformsConservative assumptions are used. The effects of tolerance values were evaluated. Those values that produce the highest doses were used in the analyses.5.1.4 Licensees should ensure that analysis assumptions and methods are compatible with the AST and the TEDE criteria. Conforms Analysis assumptions and methods are compatible with the AST and the TEDE criteria per this guidance.
B/B-UFSAR A1.183-11 REVISION 12 - DECEMBER 2008 TABLE A.1.183-1 CONFORMANCE WITH REGULATORY GUIDE 1.183 MAIN SECTIONSRG Section RG Position AnalysisComments5.3 Atmospheric dispersion values (/Q) for the EAB, the LPZ, and the control room that were approved by the staff during initial facility licensing or in subsequent licensing proceedings may be used in performing the radiological analyses identified by this guide. Methodologies that have been used for determining /Q values are documented in Regulatory Guides 1.3 and 1.4, Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," and the paper, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19". References 22 [Murphy - Campe] and 28 [RG 1.145] (of RG 1.183) should be used if the FSAR /Q values are to be revisedor if values are to be determined for new release points or receptor distances. Fumigation should be considered where applicable for the EAB and LPZ. For the EAB, the assumed fumigation period should be timed to be included in the worst 2-hour exposure period. The NRC computer code PAVAN implements Regulatory Guide 1.145 and its use is acceptable to the NRC staff. The methodology of the NRC computer code ARCON96 is generally acceptable to the NRC staff for use in determining control room /Q values.ConformsNew atmospheric dispersion values (/Q) for the EAB, the LPZ, control room, and the TSC were developed, using meteorological data for the years 1994-1998. ARCON96 and PAVAN were used with these data to determine control room, EAB, and LPZ atmospheric dispersion values. Since there is no tall stack, no fumigation is considered. Control room /Qs were developed in conformance with the guidance provided in RG 1.194.
B/B-UFSAR A1.183-12 REVISION 12 - DECEMBER 2008 TABLE A.1.183-2 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX A (LOSS-OF-COOLANT ACCIDENT)
RG Section RG Position AnalysisComments 1 Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 of this guide. Conforms (with exception relating to Table 3, footnote 11) Fission Product Inventory:
Bounding core source terms are developed using ORIGEN-2.1 based methodology. Release Fractions: Release fractions are per Table 2 of RG 1.183, and are implemented by RADTRAD. Non-LOCA Table 3 release fractions are doubled in the analyses to account for the effects of exceeding the LHGR value described in footnote 11. Timing of Release Phases:
Release Phases are per Table 4 of RG 1.183, and are implemented by RADTRAD. Radionuclide Composition:
Radionuclide grouping is per Table 5 of RG 1.183, as implemented in RADTRAD. Chemical Form: Treatment of release chemical form is per RG 1.183, Section 3.5.
2 If the sump or suppression pool pH is controlled at values of 7 or greater, the chemical form of radioiodine released to the containment should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine species, including those from iodine re-evolution, for sump or suppression pool pH values less than 7 will be evaluated on a case-by-case basis. Evaluations of pH should consider the effect of acids and bases created during the LOCA event, e.g., radiolysis products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form. ConformsThe stated distributions of iodine chemical forms are used in the analyses. The post-LOCA containment sump pH has previously been evaluated, including consideration of the effects of acids and bases created during the LOCA event, the effects of key fission product releases, and the impact of NaOH injection. Containment sump pH remains above 7 for at least 30 days.
B/B-UFSAR A1.183-13 REVISION 12 - DECEMBER 2008 TABLE A.1.183-2 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX A (LOSS-OF-COOLANT ACCIDENT)
RG Section RG Position AnalysisComments3.1 The radioactivity released from the fuel should be assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment in PWRs or the drywell in BWRs as it is released. This distribution should be adjusted if there are internal compartments that have limited ventilation exchange. The suppression pool free air volume may be included provided there is a mechanism to ensure mixing between the drywell to the wetwell. The release into the containment or drywell should be assumed to terminate at the end of the early in-vessel phase. ConformsThe radioactivity release from the fuel is assumed to instantaneously and homogeneously mix throughout the containment air space as it is released. Recirculation fans provide a mixing mechanism within the containment.
3.2 Reduction in airborne radioactivity in the containment by natural deposition within the containment may be credited. Acceptable models for removal of iodine and aerosols are described in Chapter 6.5.2, "Containment Spray as a Fission Product Cleanup System," of the Standard Review Plan (SRP), NUREG-0800 (Ref. A-1) and in NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments" (Ref. A-2). The latter model is incorporated into the analysis code RADTRAD (Ref. A-3).
ConformsThe RADTR AD computer program, including the Powers Natural Deposition algorithm based on NUREG/CR-6189, is used for modeling aerosol deposition in Containment. No natural deposition is assumed for elemental or organic iodine. The lower bound (10%) level of deposition credit is used. 3.3 Reduction in airborne radioactivity in the containment by containment spray systems that have been designed and are maintained in accordance with Chapter 6.5.2 of the SRP (Ref. A-1) may be credited. Acceptable models for the removal of iodine and aerosols are described in Chapter 6.5.2 of the SRP and NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays"1 (Ref. A-4). This simplified model is incorporated into the analysis code RADTRAD (Refs. A-1 to A-3). The evaluation of the containment sprays should address areas within the primary containment that are not covered by the spray drops. The mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified. The containment building atmosphere may be considered a single, well-mixed volume if the spray covers at least 90% of the volume and if adequate mixing of unsprayed compartments can be shown.Conforms A qualified Containment Spray System is an available design feature at both Byron and Braidwood.
The conservatively analyzed containment volume is 2.85E6 cubic feet, with 82.5% of this volume sprayed. The sprayed volume is 2.35125E6 cubic feet, unsprayed volume is 4.9875E5 cubic feet.
Transfer between these two volumes is provided by the Containment Fan Coolers. The flow rate is 65,000 cfm per fan for a total of 130,000 cfm.
B/B-UFSAR A1.183-14 REVISION 12 - DECEMBER 2008 TABLE A.1.183-2 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX A (LOSS-OF-COOLANT ACCIDENT)
RG Section RG Position AnalysisComments The SRP sets forth a maximum decontamination factor (DF) for elemental iodine based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination. The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass. There is no specified maximum DF for aerosol removal by sprays. The maximum activity to be used in determining the DF is defined as the iodine activity in the columns labeled "Total" in Tables 1 and 2 of this guide multiplied by 0.05 for elemental iodine and by 0.95 for particulate iodine (i.e., aerosol treated as particulate in SRP methodology).
It is assumed that after the end of the core activity release process the aerosols would continue to be removed at a of 6.0 hr
-1 until an overall DF of 50 is achieved. The current SRP 6.5.2 based assessment of elemental iodine removal coefficients during containment spray will continue to be used. The spray removal coefficient was determined to be 30.3 hr-1. Per SRP 6.5.2 this value is reduced to 20 hr-1. For aerosol removal the DF of 50 is reached at 2.21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />. From that point until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the removal coefficient of 0.6 hr-1 is used. For elemental iodine removal, the DF of 100 is reached at 1.926 hours0.0107 days <br />0.257 hours <br />0.00153 weeks <br />3.52343e-4 months <br />.3.4 Reduction in airborne radioactivity in the containment by in-containment recirculation filter systems may be credited if these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. A-5 and A-6). The filter media loading caused by the increased aerosol release associated with the revised source term should be addressed.
Not Applicable Not applicable for Byron or Braidwood. In-containment recirculation filters are not credited in the analyses. 3.5 Reduction in airborne radioactivity in the containment by suppression pool scrubbing in BWRs should generally not be credited. However, the staff may consider such reduction on an individual case basis. The evaluation should consider the relative timing of the blowdown and the fission product release from the fuel, the force driving the release through the pool, and the potential for any bypass of the suppression pool (Ref. 7). Analyses should consider iodine re-evolution if the suppression pool liquid pH is not maintained greater than 7. Not Applicable Not applicable for a PWR B/B-UFSAR A1.183-15 REVISION 12 - DECEMBER 2008 TABLE A.1.183-2 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX A (LOSS-OF-COOLANT ACCIDENT) RG Section RG Position AnalysisComments3.6 Reduction in airborne radioactivity in the containment by retention in ice condensers, or other engineering safety features not addressed above, should be evaluated on an individual case basis. See Section 6.5.4 of the SRP (Ref. A-1). Not Applicable Neither Byron nor Braidwood have ice condensers. No other removal mechanisms are credited other than natural deposition.3.7 The primary containment (i.e., drywell for Mark I and II containment designs) should be assumed to leak at the peak pressure Technical Specification leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% of the Technical Specification leak rate. For BWRs, leakage may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by plant configuration and analyses, to a value not less than 50% of the Technical Specification leak rate. Leakage from subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a subatmospheric condition as defined by Technical Specifications. For BWRs with Mark III containments, the leakage from the drywell into the primary containment should be based on the steaming rate of the heated reactor core, with no credit for core debris relocation. This leakage should be assumed during the two-hour period between the initial blowdown and termination of the fuel radioactivity release (gap and early in-vessel release phases). After two hours, the radioactivity is assumed to be uniformly distributed throughout the drywell and the primary containment.ConformsThe analyses follow the guidance for PWRs (the analyzed leak rate may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% of the Technical Specification leak rate). Neither Byron nor Braidwood have subatmospheric containments. 3.8 If the primary containment is routinely purged during power operations, releases via the purge system prior to containment isolation should be analyzed and the resulting doses summed with the postulated doses from other release paths. The purge release evaluation should assume that 100% of the radionuclide inventory in the reactor coolant system liquid is released to the containment at the initiation of the LOCA. This inventory should be based on the Technical Specification reactor coolant system equilibrium activity. Iodine spikes need not be considered. If the purge system is not isolated before the onset of the gap release phase, the release fractions associated with the gap release and early in-vessel phases should be considered as applicable.ConformsThe Byron and Braidwood containments can be considered to be routinely purged during power operation. Therefore, the resulting purge dose contribution is summed with the postulated doses from other release paths.
B/B-UFSAR A1.183-16 REVISION 12 - DECEMBER 2008 TABLE A.1.183-2 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX A (LOSS-OF-COOLANT ACCIDENT) RG Section RG Position AnalysisComments4.1 Leakage from the primary containment should be considered to be collected, processed by engineered safety feature (ESF) filters, if any, and released to the environment via the secondary containment exhaust system during periods in which the secondary containment has a negative pressure as defined in Technical Specifications. Credit for an elevated release should be assumed only if the point of physical release is more than two and one-half times the height of any adjacent structure. ConformsNo leakage is assumed to be collected for processing. Containment leakage is assumed to be released as a diffuse area source per RG 1.194. Since neither Byron nor Braidwood have a "tall stack," elevated releases are not assumed.4.2 Leakage from the primary containment is assumed to be released directly to the environment as a ground-level release during any period in which the secondary containment does not have a negative pressure as defined in Technical Specifications. ConformsFor EAB and LPZ doses, ground level releases are assumed. For Control Room doses, releases are based on zero-velocity vent release assumptions (ground-level equivalent).4.3 The effect of high wind speeds on the ability of the secondary containment to maintain a negative pressure should be evaluated on an individual case basis. The wind speed to be assumed is the 1-hour average value that is exceeded only 5% of the total number of hours in the data set. Ambient temperatures used in these assessments should be the 1-hour average value that is exceeded only 5% or 95% of the total numbers of hours in the data set, whichever is conservative for the intended use (e.g., if high temperatures are limiting, use those exceeded only 5%).Conforms Although Byron and Braidwood are single-containment PWRs (no secondary containments), the evaluation was performed relative to the Aux Building. The bounding 250 foot elevation wind speed exceeded only 5% of the time at Byron and Braidwood is approximately 25.2 mph. Based on representative average surface pressure coefficients for B/B-UFSAR A1.183-17 REVISION 12 - DECEMBER 2008 TABLE A.1.183-2 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX A (LOSS-OF-COOLANT ACCIDENT) RG Section RG Position AnalysisComments rectangular buildings, a wind speed of greater than 32.2 mph would be required before the TS 3.7.12 minimum negative 0.25 inches water gauge Aux Building pressure would be positive relative to outside air pressures at any building surface.4.4 Credit for dilution in the secondary containment may be allowed when adequate means to cause mixing can be demonstrated. Otherwise, the leakage from the primary containment should be assumed to be transported directly to exhaust systems without mixing. Credit for mixing, if found to be appropriate, should generally be limited to 50%. This evaluation should consider the magnitude of the containment leakage in relation to contiguous building volume or exhaust rate, the location of exhaust plenums relative to projected release locations, the recirculation ventilation systems, and internal walls and floors that impede stream flow between the release and the exhaust. N/AByron and Braidwood are PWRs with no secondary containment. 4.5 Primary containment leakage that bypasses the secondary containment should be evaluated at the bypass leak rate incorporated in the Technical Specifications. If the bypass leakage is through water, e.g., via a filled piping run that is maintained full, credit for retention of iodine and aerosols may be considered on a case-by-case basis. Similarly, deposition of aerosol radioactivity in gas-filled lines may be considered on a case-by-case basis.
N/AByron and Braidwood are PWRs with no secondary containment. Therefore, all containment leakage is released directly to the environment. 4.6 Reduction in the amount of radioactive material released from the secondary containment because of ESF filter systems may be taken into account provided that these systems meet the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref.
A-6). ConformsCredited ESF ventilation systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02.
B/B-UFSAR A1.183-18 REVISION 12 - DECEMBER 2008 TABLE A.1.183-2 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX A (LOSS-OF-COOLANT ACCIDENT) RG Section RG Position AnalysisComments5.1 With the exception of noble gases, all the fission products released from the fuel to the containment (as defined in Tables 1 and 2 of this guide) should be assumed to instantaneously and homogeneously mix in the primary containment sump water (in PWRs) or suppression pool (in BWRs) at the time of release from the core. In lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be used. Note that many of the parameters that make spray and deposition models conservative with regard to containment airborne leakage are non-conservative with regard to the buildup of sump activity. ConformsWith the exception of noble gases, all the fission products released from the fuel to the containment are assumed to instantaneously and homogeneously mix in the reactor building sump water at the time of release from the core. 5.2 The leakage should be taken as two times the sum of the simultaneous leakage from all components in the ESF recirculation systems above which the Technical Specifications, or licensee commitments to item III.D.1.1 of NUREG-0737 (Ref. A-8), would require declaring such systems inoperable. The leakage should be assumed to start at the earliest time the recirculation flow occurs in these systems and end at the latest time the releases from these systems are terminated. Consideration should also be given to design leakage through valves isolating ESF recirculation systems from tanks vented to atmosphere, e.g., emergency core cooling system (ECCS) pump miniflow return to the refueling water storage tank.
ConformsECCS leakage is analyzed at a rate twice that allowed.
ECCS leakage is a minor contributor to LOCA doses from the Byron and Braidwood plants. The accident analysis basis is 276,000 cc/hour. This leak rate is considered an upper bound that would still allow ECCS operability after 30 days, without makeup (a 12% inventory loss). 5.3 With the exception of iodine, all radioactive materials in the recirculating liquid should be assumed to be retained in the liquid phase. ConformsWith the exception of iodine, all radioactive materials in ECCS liquids are assumed to be retained in the liquid phase.
B/B-UFSAR A1.183-19 REVISION 12 - DECEMBER 2008 TABLE A.1.183-2 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX A (LOSS-OF-COOLANT ACCIDENT) RG Section RG Position AnalysisComments5.4 If the temperature of the leakage exceeds 212°F, the fraction of total iodine in the liquid that becomes airborne should be assumed equal to the fraction of the leakage that flashes to vapor. This flash fraction, FF, should be determined using a constant enthalpy, h, process, based on the maximum time-dependent temperature of the sump water circulating outside the containment:
fgf2f1 hhh FF= Where: h f1 is the enthalpy of liquid at system design temperature and pressure; h f2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212°F); and h fg is the heat of vaporization at 212°F.ConformsThe temperature of the leakage exceeds 212°F for a period of less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Therefore, a flashing factor of 10% is assumed. 5.5 If the temperature of the leakage is less than 212°F or the calculated flash fraction is less than 10%, the amount of iodine that becomes airborne should be assumed to be 10% of the total iodine activity in the leaked fluid, unless a smaller amount can be justified based on the actual sump pH history and area ventilation rates.ConformsECCS leakage flashing fractions are assumed to be 10% for the duration of the accident. 5.6 The radioiodine that is postulated to be available for release to the environment is assumed to be 97% elemental and 3% organic. Reduction in release activity by dilution or holdup within buildings, or by ESF ventilation filtration systems, may be credited where applicable. Filter systems used in these applications should be evaluated against the guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6). ConformsThe credited Control Room intake charcoal and HEPA filters meet the requirements of RG 1.52 and Generic Letter 99-02. These are credited at 95% efficiency for elemental and organic iodines. Aerosol removal efficiencies are assumed to be 99% based on the HEPA/charcoal combination.
B/B-UFSAR A1.183-20 REVISION 14 - DECEMBER 2012 TABLE A.1.183-2 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX A (LOSS-OF-COOLANT ACCIDENT)RG Section RG Position AnalysisComments The filter efficiency for the Auxiliary Building exhaust is 90.0%. These filters meet the requirements of RG 1.52 and Generic Letter 99-027.0 The radiological consequences from post-LOCA primary containment purging as a combustible gas or pressure control measure should be analyzed. If the installed containment purging capabilities are maintained for purposes of severe accident management and are not credited in any design basis analysis, radiological consequences need not be evaluated. If the primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref.
A-6).Conforms Although the UFSAR discusses containment purge for hydrogen control, this will not be considered as a release pathway during a LOCA. The system that would be used for hydrogen purge is not safety grade and thus does not meet the system operating requirements for a design basis accident.
B/B-UFSAR A1.183-21 REVISION 12 - DECEMBER 2008 TABLE A.1.183-3 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX B (FUEL HANDLING ACCIDENT) RG Section RG Position AnalysisComments 1 Acceptable assumptions regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 of this guide.
Conforms See Table A.1.183-1 for conformance with Regulatory Guide 1.183 Appendix B (Fuel Handling Accident). The bounding inventory of fission products in the reactor core and available for release to the containment is based on the maximum full power operation of the core with current licensed values for fuel enrichment, fuel burnup, and a core power equal to the current licensed rated thermal power times the ECCS evaluation uncertainty. Additional conservatisms are added as discussed in Table A.1.183-1 of this document to ensure the bounding source term was determined.1.1 The number of fuel rods damaged during the accident should be based on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load or the weight of a dropped fuel assembly (plus any attached handling grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g., events over the reactor vessel), should be considered. ConformsThe number of fuel rods damaged is equal to one fuel assembly.
As currently described in UFSAR Section 15.7.4 (Fuel Handling Accidents), the accident is defined as the drop of a spent fuel assembly (SFA) onto the spent fuel pool floor or the core, resulting in the postulated rupture of the cladding of all fuel rods in one assembly.
B/B-UFSAR A1.183-22 REVISION 12 - DECEMBER 2008 TABLE A.1.183-3 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX B (FUEL HANDLING ACCIDENT) RG Section RG Position AnalysisComments1.2 The fission product release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. All the gap activity in the damaged rods is assumed to be instantaneously released. Radionuclides that should be considered include xenons, kryptons, halogens, cesiums, and rubidiums.
Exception taken(al-ternative treatment used) Since several fuel assemblies exceed the guidance outlined in Footnote 11, the gap release fractions are doubled for conservatism. This treatment (previously approved for Fort Calhoun is conservative as previously discussed in Table A of this Compliance Table (Item 3.2).1.3 The chemical form of radioiodine released from the fuel to the spent fuel pool should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The CsI released from the fuel is assumed to completely dissociate in the pool water. Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. This is assumed to occur instantaneously. The NRC staff will consider, on a case-by-case basis, justifiable mechanistic treatment of the iodine release from the pool. Conforms All iodine added to the reactor vessel or spent fuel pool is assumed to instantaneously dissociate and re-evolve as elemental iodine and treated appropriately with regard to pool pH and is assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide.2 If the depth of water above the damaged fuel is 23 feet or greater, the decontamination factors for the elemental and organic species are 500 and 1, respectively, giving an overall effective decontamination factor of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water). This difference in decontamination factors for elemental (99.85%) and organic iodine (0.15%) species results in the iodine above the water being composed of 57% elemental and 43% organic species. If the depth of water is not 23 feet, the decontamination factor will have to be determined on a case-by-case method (Ref. B-1)ConformsThe analyzed water depthabove damaged fuel is 23 feet. This value corresponds to the minimum depth of water coverage over the top of irradiated fuel assemblies seated in the spent fuel pool racks within the spent fuel pool, as per TS 3.7.14.
B/B-UFSAR A1.183-23 REVISION 12 - DECEMBER 2008 TABLE A.1.183-3 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX B (FUEL HANDLING ACCIDENT) RG Section RG Position AnalysisComments Therefore, an overall DF of 200 is used per this guidance.
Iodine above the water is assumed to be composed of 57% elemental and 43% organic species 3 The retention of noble gases in the water in the fuel pool or reactor cavity is negligible (i.e., decontamination factor of 1). Particulate radionuclides are assumed to be retained by the water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).ConformsDF = 1 for noble gas isotopes; DF = infinite for particulate radionuclides. 4.1 The radioactive material that escapes from the fuel pool to the fuel building is assumed to be released to the environment over a 2-hour time period.ConformsThe release is assumed to occur over a two-hour period.4.2 A reduction in the amount of radioactive material released from the fuel pool by engineered safety feature (ESF) filter systems may be taken into account provided these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2, B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.Conforms All ESF filtration systems credited in the analyses are qualified in accordance with the references cited in this section. 4.3 The radioactivity release from the fuel pool should be assumed to be drawn into the ESF filtration system without mixing or dilution in the fuel building. If mixing can be demonstrated, credit for mixing and dilution may be considered on a case-by-case basis. This evaluation should consider the magnitude of the building volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the pool, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the pool and the exhaust plenums.Conforms As per RG 1.183, the release from the fuel building to the environment is assumed over a 2-hour time period. To assure this, the refueling floor exhaust rate is set artificially high at 5 times this value or 0.118 air changes per minute during Control Room emergency mode of operation.
B/B-UFSAR A1.183-24 REVISION 12 - DECEMBER 2008 TABLE A.1.183-3 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX B (FUEL HANDLING ACCIDENT) RG Section RG Position AnalysisComments5.1 If the containment is isolated during fuel handling operations, no radiological consequences need to be analyzed.
Not Applicable Containment isolation is not credited in the analysis.5.2 If the containment is open during fuel handling operations, but designed to automatically isolate in the event of a fuel handling accident, the release duration should be based on delays in radiation detection and completion of containment isolation. If it can be shown that containment isolation occurs before radioactivity is released to the environment, no radiological consequences need to be analyzed.Conforms Automatic containment isolation is not credited. Therefore, a radiological consequence analysis is performed. 5.3 If the containment is open during fuel handling operations (e.g., personnel air lock or equipment hatch is open), the radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2-hour time period.
Note 3: The staff will generally require that Technical Specifications allowing such operations include administrative controls to close the airlock, hatch, or penetrations within 30 minutes. Such administrative controls will generally require that a dedicated individual be present, with the necessary equipment available, to restore containment closure should a fuel handling accident occur. Radiological analyses should generally not credit this manual isolation.
Conforms (with site-specific exceptions as noted in Attachment 5 of the submittal) The radioactive material that escapes from the reactor cavity pool to the containment is released to the environment over a 2-hour time period. 5.4 A reduction in the amount of radioactive material released from the containment by ESF filter systems may be taken into account provided that these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses. ConformsFor non-Recently Irradiated Fuel, no filtration of the radioactive gas released from the pool or automatic isolation of the accident location is assumed, with essentially all of the activity reaching the refueling floor airspace exhausted to the environment within two hours after the accident.
B/B-UFSAR A1.183-25 REVISION 12 - DECEMBER 2008 TABLE A.1.183-3 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX B (FUEL HANDLING ACCIDENT) RG Section RG Position AnalysisComments For Recently Irradiated Fuel, an additional FHA analysis was performed with containment closure established or with the FHB ventilation system operable. The results of this analysis also met the limits of 10 CFR 50.67 assuming a minimum decay time of six hours. The six-hour minimum decay time is inconsequential as it is physically impossible to remove the reactor head and move fuel within the first six hours after the reactor is subcritical.5.5 Credit for dilution or mixing of the activity released from the reactor cavity by natural or forced convection inside the containment may be considered on a case-by-case basis. Such credit is generally limited to 50% of the containment free volume. This evaluation should consider the magnitude of the containment volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the reactor cavity, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums. ConformsThe activity is instantaneously released from the fuel into the containment and is assumed to mix with 100% of the containment volume to calculate a hypothetical release rate with which to remove nearly all the activity within a two-hour period. This creates a conservative release rate over the two-hour release period.
B/B-UFSAR A1.183-26 REVISION 12 - DECEMBER 2008 TABLE A.1.183-4 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX H (PWR ROD EJECTION ACCIDENT) RG Section RG Position AnalysisComments 1 Assumptions acceptable to the NRC staff regarding core inventory are in Regulatory Position 3 of this guide. For the rod ejection accident, the release from the breached fuel is based on the estimate of the number of fuel rods breached and the assumption that 10% of the core inventory of the noble gases and iodines is in the fuel gap. The release attributed to fuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting and the assumption that 100% of the noble gases and 25% of the iodines contained in that fraction are available for release from containment. For the secondary system release pathway, 100% of the noble gases and 50% of the iodines in that fraction are released to the reactor coolant. ConformsTheCREA core source terms are those associated with a DBA power level of 3658.3 MWth, which includes an additional 2% power over that of the full licensed power to account for uncertainty.
The sudden rod ejection and localized temperature spike associated with the CREA results in the damage of 10% of the core. Only 2.5 % of the damaged core releases melted fuel activity, i.e., 0.00250 of the total core melts. Therefore, the source term available for release is associated with this fraction of melted fuel and the fraction of core activity existing in the gap. A peaking factor of 1.7 is also applied.
2 If no fuel damage is postulated for the limiting event, a radiological analysis is not required as the consequences of this event are bounded by the consequences projected for the loss-of-coolant accident (LOCA), main steam line break, and steam generator tube rupture.
Not Applicable Since fuel damage is postulated, a radiological consequence analysis is performed.
B/B-UFSAR A1.183-27 REVISION 12 - DECEMBER 2008 TABLE A.1.183-4 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX H (PWR ROD EJECTION ACCIDENT) RG Section RG Position AnalysisComments 3 Two release cases are to be considered. In the first, 100% of the activity released from the fuel should be assumed to be released instantaneously and homogeneously through the containment atmosphere. In the second, 100% of the activity released from the fuel should be assumed to be completely dissolved in the primary coolant and available for release to the secondary system. ConformsFor Case 1, the ejectedcontrol rod is assumed to breach the reactor pressure vessel (RPV), effectively causing the equivalent of a small break loss of coolant accident. In this case, all activity from damaged fuel that has been mixed with the primary coolant of the Reactor Coolant System (RCS) leaks directly to the containment volume. This flashed release is assumed to instantaneously and homogeneously mix with the containment atmosphere, and is available for release to the environment via a Containment leak rate limit, or L
- a. For Case 2, no breach of the RPV is assumed following the rod ejection. In this case, RCS integrity is maintained and all activity from damaged fuel that has been mixed with the RCS leaks to the secondary side coolant through the Steam Generator (SG) tubes via the Tech. Spec. primary to secondary coolant leakage rate of 1.0 gpm.
B/B-UFSAR A1.183-28 REVISION 12 - DECEMBER 2008 TABLE A.1.183-4 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX H (PWR ROD EJECTION ACCIDENT) RG Section RG Position AnalysisComments From here, activity is available for release to the environment by steaming of the SG Power-Operated Relief Valves (PORVs). In addition to the activity released from the primary to secondary coolant, pre-existing Tech. Spec. iodine activity in the secondary coolant system is assu med to also be released.
4 The chemical form of radioiodine released to the containment atmosphere should be assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. If containment sprays do not actuate or are terminated prior to accumulating sump water, or if the containment sump pH is not controlled at values of 7 or greater, the iodine species should be evaluated on an individual case basis. Evaluations of pH should consider the effect of acids created during the rod ejection accident event, e.g., pyrolysis and radiolysis products. With the exception of elemental and organic iodine and noble gases, fission products should be assumed to be in particulate form. Conforms All iodine released from the SGs is conservatively assumed to be of the elemental species. This is done for RADTRAD simulation considerations, and is consistent with the RG 1.183, because elemental and organic iodine are identically treated by the computer model. 5 Iodine releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic. Conforms All iodine released from the SGs is assumed to be of the elemental species. This is done for RADTRAD simulation considerations, and is consistent with the RG 1.183 specification of 97% elemental and 3% organic, because elemental and organic iodine are identically treated by the computer model. 6 Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material in and from the containment are as follows.Conforms(See sections 6.1 and 6.2)
B/B-UFSAR A1.183-29 REVISION 12 - DECEMBER 2008 TABLE A.1.183-4 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX H (PWR ROD EJECTION ACCIDENT) RG Section RG Position AnalysisComments6.1 A reduction in the amount of radioactive material available for leakage from t he containment that is due to natural deposition, containment sprays, recirculating filter systems, dual containments, or other engineered safety feat ures may be taken into account. Refer to Appendix A to this guide for guidance on acceptable methods and as sumptions for evaluating these mechanisms.
ConformsThe RADTRAD computer program, including the Powers Natural Deposition algorithm based on NUREG/CR-6189, is used for modeling aerosol deposition in Containment. No natural deposition is assumed for elemental or organic iodine. The lower bound (10%) level of deposition credit is used.
Decay of radioactivity is credited in all compartments, prior to release. This is implemented in RADTRAD using the half-lives in the Nuclide Inventory File (NIF). The RADTRAD decay plus daughter option is used. In reality, daughter products such as xenon from iodines or iodines from tellurium are unlikely to readily escape from the fuel matrix in which the parent iodine or tellurium is contained. Nevertheless, the RADTRAD feature to include daughter effects is selected for conservatism.
No credit for containment spray is taken.
B/B-UFSAR A1.183-30 REVISION 12 - DECEMBER 2008 TABLE A.1.183-4 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX H (PWR ROD EJECTION ACCIDENT) RG Section RG Position AnalysisComments6.2 The containment should be assumed to leak at the leak rate incorporated in the Technical Specifications at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate for the remaining duration of the accident. Peak accident pressure is the maximum pressure defined in the Technical Specifications for containment leak testing. Leakage from subatmospheric containments is assumed to be terminated when the containment is brought to a subatmospheric condition as defined in Technical Specifications. ConformsThe containment is assumed to leak at the leak rate incorporated in the Technical Specifications at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate for the remaining duration of the accident. 7.1 A leak rate equivalent to the primary-to-secondary leak rate Limiting Condition for Operation specified in the Technical Specifications should be assumed to exist until shutdown cooling is in operation and releases from the steam generators have been terminated. ConformsThe leak rate equivalent to the primary-to-secondary leak rate Limiting Condition for Operation specified in the Technical Specifications is assumed to exist until shutdown cooling is in operation and releases from the steam generators have been terminated.7.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate Technical Specifications. These tests typically are based on cooled liquid. The facility's instrumentation used to determine leakage typically is located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).
ConformsThe density is assumed to be 1.0 gm/cc (62.4 lbm/ft3) 7.3 All noble gas radionuclides released to the secondary system are assumed to be released to the environment without reduction or mitigation.ConformsNoble gases are released without reduction or mitigation. 7.4 The transport model described in assumptions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulates. ConformsThe transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E was utilized for iodine and particulates.
B/B-UFSAR A1.183-31 REVISION 12 - DECEMBER 2008 TABLE A.1.183-5 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX G (PWR LOCKED ROTOR ACCIDENT) RG Section RG Position AnalysisComments 1 Assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel are in Regulatory Position 3 of this regulatory guide. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.
A conserva-tive exception is taken regarding release fractions The analysis does not fully comply with Note 11 of Table 3 since typical Byron and Braidwood core designs indicate that there are fuel assemblies that exceed the 6.3 kW/ft while >54GWD/MTU. Previous analyses (ANS 5.4) for TMI-1 have shown that those fuel assemblies exceeding these limits had no increase in gap release fractions of concern. Therefore, doubling of the "Other Noble Gases', "Other Halogens", and "Alkali Metals" gap fractions in Table 3 is conservative as used and approved in the Fort Calhoun AST submittal.
Additionally, a peaking factor of 1.7 is used for DBA events that do not involve the entire core.
2 If no fuel damage is postulated for the limiting event, a radiological analysis is not required as the consequences of this event are bounded by the consequences projected for the main steam line break outside containment. Not ApplicableFuel damage is assumed. Therefore, a specific analysis is performed.
3 The activity released from the fuel should be assumed to be released instantaneously and homogeneously through the primary coolant.
ConformsThe activity is assumed to be released instantaneously and homogeneously through the primary coolant.
B/B-UFSAR A1.183-32 REVISION 12 - DECEMBER 2008 TABLE A.1.183-5 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX G (PWR LOCKED ROTOR ACCIDENT) RG Section RG Position AnalysisComments 4 The chemical form of radioiodine released from the fuel should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during normal operations, including iodine spiking.ConformsIodinechemical form is in accordance with this guidance (97% elemental, 3% organic iodines). 5.1 The primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate Limiting Condition for Operation specified in the Technical Specifications. The leakage should be apportioned between the steam generators in such a manner that the calculated dose is maximized.
ConformsNeither Byron nor Braidwood have implemented alternative repair criteria. Therefore, the primary-to-secondary leak rate in the steam generators is assumed to be the leak rate Limiting Condition for Operation specified in the Technical Specifications.
The design basis leak rate is 0.218 gpm per intact SG, totaling 0.654 gpm. 5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate Technical Specifications. These tests are typically based on cool liquid. Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).ConformsThe density is assumed to be 1.0 gm/cc (62.4 lbm/ft3)
B/B-UFSAR A1.183-33 REVISION 12 - DECEMBER 2008 TABLE A.1.183-5 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX G (PWRLOCKED ROTOR ACCIDENT)RG Section RG Position AnalysisComments5.3 The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100°C (212° F). The release of radioactivity should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.ConformsThe steaming releaseand primary-to-secondary coolant leakage is postulated to end at 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, when the RCS and secondary loop have equilibrated.5.4 The release of fission products from the secondary system should be evaluated with the assumption of a coincident loss of offsite power. Conforms A coincident loss of offsite power is assumed. 5.5 All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.ConformsNoble gases are released without reduction or mitigation. 5.6 The transport model described in assumptions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulates.
ConformsThe transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E was utilized for iodine and particulates.
B/B-UFSAR A1.183-34 REVISION 12 - DECEMBER 2008 TABLE A.1.183-6 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX E (PWR MAIN STEAM LINE BREAK) RG Section RG Position AnalysisComments 1 Assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 of this regulatory guide. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. The fuel damage estimate should assume that the highest worth control rod is stuck at its fully withdrawn position. ConformsNo fuel damage is postulated to occur during the MSLB (see Section 2 below). 2 If no or minimal 2 fuel damage is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by the Technical Specifications. Two cases of iodine spiking should be assumed.
Footnote 2: The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum Technical Specification values, whichever maximizes the radiological consequences. In determining dose equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.ConformsThe activity assumed in the analysis is based on the activity associated with the maximum Technical Specification values. In determining dose equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes is included. 2.1 A reactor transient has occurred prior to the postulated main steam line break (MSLB) and has raised the primary coolant iodine concentration to the maximum value (typically 60 µCi/gm DE I-131) permitted by the Technical Specifications (i.e., a pre-accident iodine spike case).
ConformsThis analyzed case involves a 60
µCi/gm pre-accident Iodine spike, consistent with the Technical Specification operational Reactor Coolant System (RCS) activity concentration limit for an assumed spike. All of the spike activity is homogeneously mixed in the primary coolant, prior to accident initiation.
B/B-UFSAR A1.183-35 REVISION 12 - DECEMBER 2008 TABLE A.1.183-6 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX E (PWR MAIN STEAM LINE BREAK) RG Section RG Position AnalysisComments2.2 The primary system transient associated with the MSLB causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 500 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 µCi/gm DE I-131) specified in Technical Specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel gap of all fuel pins. Conforms(Iodine spike has been determined to last for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> instead of 8) This case involves an accident initiated iodine spike that occurs concurrently with the release of fluid from the primary and secondary coolant systems. This spike results in a release rate from the operating limit defective fuel fraction that is 500 times the normal rate. Conservative Byron and Braidwood analyses have shown that after 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> the total iodine gap activity of the defective fuel will have been completely released into the primary coolant.
3 The activity released from the fuel should be assumed to be released instantaneously and homogeneously through the primary coolant. ConformsThe released activity is assumed to be dispersed instantaneously and homogeneously through the primary coolant.
4 The chemical form of radioiodine released from the fuel should be assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine released during normal operations, including iodine spiking. ConformsIodine chemical form is in accordance with this guidance (i.e., 97% elemental and 3% organic).
B/B-UFSAR A1.183-36 REVISION 12 - DECEMBER 2008 TABLE A.1.183-6 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX E (PWR MAIN STEAM LINE BREAK) RG Section RG Position AnalysisComments5.1 For facilities that have not implemented alternative repair criteria (see Ref. E-1, DG-1074), the primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate Limiting Condition for Operation specified in the Technical Specifications. For facilities with traditional generator specifications (both per generator and total of all generators), the leakage should be apportioned between affected and unaffected steam generators in such a manner that the calculated dose is maximized.
ConformsNeither Byron nor Braidwood have implemented alternative repair criteria. Therefore, the primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate Limiting Condition for Operation specified in the Technical Specifications. Activity that originates in the primary RCS is released to the secondary coolant by means of the primary-to-secondary coolant leak rate. This design basis leak rate value is 0.218 gpm, per intact SG, totaling 0.654 gpm, and 0.5 gpm for the faulted SG with the broken steam line.5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should be consistent with the basis of the parameter being converted. The ARC leak rate correlations are generally based on the collection of cooled liquid. Surveillance tests and facility instrumentation used to show compliance with leak rate Technical Specifications are typically based on cooled liquid. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).ConformsThe density is assumed to be 1.0 gm/cc (62.4 lbm/ft3). 5.3 The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100°C (212°F). The release of radioactivity from unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated. ConformsThe steaming release and primary-to-secondary coolant leakage is postulated to end at 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, when the RCS and secondary loop have equilibrated.
B/B-UFSAR A1.183-37 REVISION 12 - DECEMBER 2008 TABLE A.1.183-6 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX E (PWR MAIN STEAM LINE BREAK) RG Section RG Position AnalysisComments5.4 All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.ConformsNoble gases are released without reduction or mitigation.5.5 The transport model described in this section should be utilized for iodine and particulate releases from the steam generators. This model is shown in Figure E-1 and summarized below:
ConformsThe transport model described in this section is utilized for iodine and particulate releases from the steam generators. 5.5.1 A portion of the primary-to-secondary leakage will flash to vapor, based on the thermodynamic conditions in the reactor and secondary coolant. During periods of steam generator dryout, all of the primary-to-secondary leakage is assumed to flash to vapor and be released to the environment with no mitigation.
With regard to the unaffected steam generators used for plant cooldown, the primary-to-secondary leakage can be assumed to mix with the secondary water without flashing during periods of total tube submergence. Primary to secondary coolant leakage through the faulted steam generator conservatively goes directly to the environment, without mixing with any secondary coolant. Therefore, under the assumed dry-out conditions, no partitioning of any nuclides is expected to occur in this release pathway. For all post-accident releases through the
B/B-UFSAR A1.183-38 REVISION 12 - DECEMBER 2008 TABLE A.1.183-6 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX E (PWR MAIN STEAM LINE BREAK) RG Section RG Position AnalysisComments PORVs of the intact SG loops, the mechanism for release to the environment is steaming of the secondary coolant. Because of this release dynamic, a reduction is taken in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water. For Iodine, the partitioning factor of 0.01 was taken directly from RG 1.183. Reviewing the specified AST release fractions, it is concluded that the only nuclides other than iodines to be released from the core source term are noble gas nuclides. Because of the volatility of noble gases, no partitioning is assumed for any such isotopes.5.5.2 The leakage that immediately flashes to vapor will rise through the bulk water of the steam generator and enter the steam space. Credit may be taken for scrubbing in the generator, using the models in NUREG-0409, "Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident" (Ref. E-2), during periods of total submergence of the tubes.ConformsSee Comments for Section 5.5.1. 5.5.3 The leakage that does not immediately flash is assumed to mix with the bulk water.ConformsSee Comments for Section 5.5.1.
B/B-UFSAR A1.183-39 REVISION 12 - DECEMBER 2008 TABLE A.1.183-6 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX E (PWR MAIN STEAM LINE BREAK) RG Section RG Position AnalysisComments5.5.4 The radioactivity in the bulk water is assumed to become vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine of 100 may be assumed. The retention of particulate radionuclides in the steam generators is limited by the moisture carryover from the steam generators.ConformsThe specified partition coefficient is used in the analysis. 5.6 Operating experience and analyses have shown that for some steam generator designs, tube uncovery may occur for a short period following any reactor trip (Ref. E-3). The potential impact of tube uncovery on the transport model parameters (e.g., flash fraction, scrubbing credit) needs to be considered. The impact of emergency operating procedure restoration strategies on steam generator water levels should be evaluated. ConformsSee Comments for Section 5.5.1.
B/B-UFSAR A1.183-40 REVISION 15 - DECEMBER 2014 TABLE A.1.183-7 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX F (PWR STEAM GENERATOR TUBE RUPTURE ACCIDENT) RG Section RG Position AnalysisComments 1 Assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel are in Regulatory Position 3 of this guide. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.
ConformsSee Table A.1.183-1 for conformance with Regulatory Guide 1.183 Appendix F. The bounding inventory of fission products in the reactor core and available for release to the containment is based on the maximum full power operation of the core with current licensed values for fuel enrichment, fuel burnup, and a core power equal to the current licensed rated thermal power times the ECCS evaluation uncertainty. Additional conservatisms are added as discussed in Table A.1.183-1 of this document to ensure the bounding source term was determined.2 If no or minimal 2 fuel damage is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by Technical Specification. Two cases of iodine spiking should be assumed. Footnote #2: The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum Technical Specification values, whichever maximizes the radiological consequences. In determining dose equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included. ConformsThe design basis assumes no fuel damage for the postulated SGTR event. For this SGTR accident, the source terms are defined by the Technical Specification activity release rates from a maximum failed fuel fraction assumed during operation, which are characterized by the equilibrium 1.0 Ci/gm Dose Equivalent (DE) I-131 iodine activity concentration in the primary reactor coolant system. The noble gas inventory in the RCS is based on operation at the Technical Specification limit of 603 micro-Ci/gm DE Xe-133.
B/B-UFSAR A1.183-41 REVISION 15 - DECEMBER 2014 TABLE A.1.183-7 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX F (PWR STEAM GENERATOR TUBE RUPTURE ACCIDENT) RG Section RG Position AnalysisComments Because no fuel damage is assumed for this accident, only iodine and noble gas isotopes are modeled to contribute to dose. To identify the worst-case SGTR accident, however, two different cases of iodine spiking are analyzed, per regulatory guidance (Pre-Accident Iodine Spike and Concurrent Iodine Spike).2.1 A reactor transient has occurred prior to the postulated steam generator tube rupture (SGTR) and has raised the primary coolant iodine concentration to the maximum value (typically 60 µCi/gm DE I-131) permitted by the Technical Specifications (i.e., a pre-accident iodine spike case). ConformsThis analyzed case involves a 60 µCi/gm pre-accident Iodine spike, consistent with the Technical Specification operational Reactor Coolant System (RCS) activity concentration limit for an assumed spike. All of the spike activity is homogeneously mixed in the primary coolant, prior to accident initiation.2.2 The primary system transient associated with the SGTR causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 µCi/gm DE I-131) specified in Technical Specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel gap of all fuel pins. ConformsThe second analyzed case involves an accident initiated iodine spike that occurs concurrently with the release of fluid from the primary and secondary coolant systems. This spike results in a release rate from defective fuel that is 335 times the normal rate, and lasts for an 8-hour duration.
B/B-UFSAR A1.183-42 REVISION 12 - DECEMBER 2008 TABLE A.1.183-7 CONFORMANCE WITH REGULATORY GUIDE 1.183 APPENDIX F (PWR STEAM GENERATOR TUBE RUPTURE ACCIDENT) RG Section RG Position AnalysisComments 3 The activity released from the fuel, if any, should be assumed to be released instantaneously and homogeneously through the primary coolant.ConformsMixing in the primary coolant is assumed to be instantly and homogeneously.
4 Iodine releases from the steam generators to the environment should be assumed to be 97% elemental and 3% organic. ConformsSuch iodine releases are assumed to be 97% elemental and 3% organic.5.1 The primary-to-secondary leak rate in the steam generators should be assumed to be the leak rate Limiting Condition for Operation specified in the Technical Specifications. The leakage should be apportioned between affected and unaffected steam generators in such a manner that the calculated dose is maximized. Conforms Activity that originates in the primary RCS is released to the secondary coolant by means of the primary-to-secondary coolant leak rate. This design basis leak rate value is 0.218 gpm per intact SG, totaling 0.654 gpm.5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g., lbm/hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate Technical Specifications. These tests are typically based on cool liquid. Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).ConformsThe density is assumed to be 1.0 gm/cc (62.4 lbm/ft3) 5.3 The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100°C (212° F). The release of radioactivity from the unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated. ConformsRelease of activity terminates when shutdown cooling has been established. 5.4 The release of fission products from the secondary system should be evaluated with the assumption of a coincident loss of offsite power. Conforms A coincident loss of offsite power is assumed. 5.5 All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.ConformsNoble gases are released without reduction or mitigation. 5.6 The transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E should be utilized for iodine and particulates. ConformsThe transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E was utilized for iodine and particulates.
B/B-UFSAR A1.194-1 REVI SION 12 - DECEMBER 2008
Regulatory Guide 1.194 Atmospheric Relative C oncentrations For Control Room Radiological Habitability Assessments At Nuclear Power Plants The Licensee complies with Revision 0 of the regulatory position with comments and exceptions as listed in UFSAR Table A1.194-1.
B/B-UFSAR A1.194-2 REVISION 12 - DE CEMBER 2008 TABLE A.1.194-1 CONFORMANCE WITH REGULATORY GUIDE 1.194 (DIFFUSE AREA SOURCE GUIDANCE) RG Section RG Position AnalysisComments3.2.4 Examples of possible area sources are postulated releases from the surface of a reactor or a secondary containment building.
A reasonable approach (is) to model the building surface as a vertical planar area source. This approach is not intended to address dispersion resulting from building-induced turbulence. Treatment of a release as a diffuse source will be acceptable for design basis calculations if the guidance herein is followed.ConformsIntroductory information excerpted - no requirements. 3.2.4.1 Diffuse source modeling should be used only for those situations in which the activity being released is homogeneously distributed throughout the building and when the assumed release rate from the building surface would be reasonably constant over the surface of the building. For example, steam releases within a turbine building with roof ventilators or louvered walls would generally not be suitable for modeling as a diffuse source. (See Regulatory Positions 3.2.4.7 and 3.2.4.8.)ConformsUsed only for containment building, where the situation would comply with this guidance. 3.2.4.2 Since leakage is more likely to occur at a penetration, analysts must consider the potential impact of building penetrations exposed to the environment* within this modeled area. If the penetration release would be more limiting, the diffuse area source model should not be used. Releases from personnel air locks and equipment hatches exposed to the environment, or containment purge releases prior to containment isolation, may need to be treated differently. It may be necessary to consider several cases to ensure that the /Q value for the most limiting location is identified. *Penetrations that are enclosed within safety-related structures need not be considered in this evaluation if the release would be captured and released via a plant ventilation system, as ventilation system releases should have already been addressed as a separate release point. ConformsContainment radioactivity releases through penetrations and personnel/equipment hatches into the auxiliary building are served by otherwise un-credited HEPA filters and charcoal adsorbers before release through the plant vent. This filtration more than offsets differences between plant vent and containment diffuse area source /Qs. Therefore, unfiltered containment diffuse area treatment can be conservatively applied to these penetrations. Leakage through the secondary personnel/equipment hatch would be unfiltered, but the hatch is located on the far
B/B-UFSAR A1.194-3 REVISION 12 - DE CEMBER 2008 TABLE A.1.194-1 CONFORMANCE WITH REGULATORY GUIDE 1.194 (DIFFUSE AREA SOURCE GUIDANCE) RG Section RG Position AnalysisComments side of the containment buildings with respect to the Control Room intakes, with /Qs more favorable than the containment building diffuse source /Qs. Containment vent penetrations are exhausted through the plant vent, but such leakage, if not directly into the auxiliary building, would be past miniflow or normal ventilation HEPA filters, or the post-LOCA purge HEPA filters and charcoal adsorbers. Therefore, unfiltered containment diffuse area treatment can be conservatively applied to these penetrations.
Containment purge supply penetration leakage, if not directly into the auxiliary building, would be to the auxiliary building supply air intake such that it would be drawn back into the auxiliary building. Therefore, unfiltered containment diffuse area treatment can be conservatively applied to these penetrations.
B/B-UFSAR A1.194-4 REVISION 12 - DE CEMBER 2008 TABLE A.1.194-1 CONFORMANCE WITH REGULATORY GUIDE 1.194 (DIFFUSE AREA SOURCE GUIDANCE) RG Section RG Position AnalysisComments Penetration leakage into the steam tunnel is not from the containment atmosphere, due to the barrier provided by the steam generators.3.2.4.3 The total release rate (e.g., Ci/second) from the building atmosphere is to be used in conjunction with the diffuse area source /Q in assessments. This release rate is assumed to be equally distributed over the entire diffuse source area from which the radioactivity release can enter the environment. For freestanding containments, this would be the entire periphery above grade or above a building that surrounds the lower elevations of the containment. When a licensee can justify assuming collection of a portion of the release from the containment within the surrounding building, the total release from the containment may be apportioned between the exposed and enclosed building surfaces. Similarly, if the building atmosphere release is modeled through more than one simultaneous pathway (e.g., drywell leakage and main steam safety valve leakage in a BWR), only that portion of the total release released through the building surface should be used with the diffuse area /Q. ~The release rate should not be averaged or otherwise apportioned over the surface area of the building. For example, reducing the release rate by 50 percent because only 50 percent of the surface faces the control room intake would be inappropriate.ConformsThe containment buildings are freestanding containments, so the source area is the entire periphery above grade. 3.2.4.4 ARCON96 uses two initial diffusion coefficients entered by the user to represent the area source. There are insufficient field measurements to mechanistically model these initial diffusion coefficients. The following deterministic equations should be used in the absence of site-specific empirical data.*
Conforms ARCON96 and the two equations are utilized.
B/B-UFSAR A1.194-5 REVISION 12 - DE CEMBER 2008 TABLE A.1.194-1 CONFORMANCE WITH REGULATORY GUIDE 1.194 (DIFFUSE AREA SOURCE GUIDANCE) RG Section RG Position AnalysisComments Sigma Y o = Width source area
6 Sigma Z o = Height source area 6 *See Regulatory Position 7 regarding the use of site-specific empirical measurements.3.2.4.5 The height and width of the area source (e.g., the building surface) are taken as the maximum vertical and horizontal dimensions of the above-grade building cross-sectional area perpendicular to the line of sight from the building center to the control room intake (see Figure 2). These dimensions are projected onto a vertical plane perpendicular to the line of sight and located at the closest point on the building surface to the control room intake. The release height is set at the vertical center of the projected plane. The source-to-receptor distance (slant path) is measured from this point to the control room intake.Conforms3.2.4.6 Intentional releases from a secondary containment (e.g., standby gas treatment systems (SGTS) at BWR reactors) or annulus ventilation systems in dual containment structures should be treated as a ground-level release or an elevated stack release, as appropriate. The diffuse area source model may be appropriate for time intervals for which the secondary containment or annulus ventilation system is not capable of maintaining the requisite negative pressure differential specified in Technical Specifications or in the UFSAR. Secondary containment bypass leakage (i.e., leakage from the primary containment that bypasses the secondary containment and is not collected by the SGTS) should be treated as a ground-level release or an elevated stack release, as appropriate.
Not applicable 3.2.4.7 A second possible application of the diffuse area source model is determining a /Q value for multiple (i.e., 3 or more) roof vents. This treatment would be appropriate for configurations in which (1) the vents are in a close arrangement, (2) no individual vent is significantly* closer to the control room intake than the center of the area source, (3) the release rate from each vent is approximately the same, and (4) no credit is taken for plume rise. The Not applicable
B/B-UFSAR A1.194-6 REVISION 12 - DE CEMBER 2008 TABLE A.1.194-1 CONFORMANCE WITH REGULATORY GUIDE 1.194 (DIFFUSE AREA SOURCE GUIDANCE) RG Section RG Position AnalysisComments distance to the receptor is measured from the closest point on the perimeter of the assumed area source. For assumed areas that are not circular, the area width is measured perpendicular to the line of sight from the center of the assumed source to the control room intake. The initial diffusion coefficient sigma Yo is found by Equation 3; sigma Zo is assumed to be 0.0.
- The degree of significance will depend on the radius or width of the assumed area and the proximity of the vent cluster to the control room intake. As the radius decreases or the distance from the cluster to the control room intake increases, the less significance the position of any one vent has. 3.2.4.8 A third possible application of the diffuse area source model is determining a /Q value for large louvered panels or large openings (e.g., railway doors on BWR Mark I plants) on vertical walls. This treatment would be appropriate for a louvered panel or opening when (1) the release rate from the building interior is essentially equally dispersed over the entire surface of the panel or opening and (2) assumptions of mixing, dilution, and transport within the building necessary to meet condition 1 are supported by the interior building arrangement. The staff has traditionally not allowed credit for mixing and holdup in turbine buildings because of the buoyant nature of steam releases and the typical presence of high volume roof exhaust ventilators. The distance to the receptor and the release height is measured from the center of the louvered panel or opening. Initial diffusion coefficients are found using Equations 3 and 4 assuming the width and height is that of the panel or opening rather than that of the building. If the area source and the intake are on the same building surface such that wind flows along the building surface would transport the release to the intake, the initial dispersion coefficient will need to be adjusted. If the included angle between the source-receptor line of sight and the vertical axis of the assumed source is less than 45 degrees, sigma Y o should be set to 0.0. If the included angle between the source receptor line of sight and the horizontal axis of the assumed source is less than 45 degrees, sigma Z o should be set to 0.0.
Not applicable B/B-UFSAR A1.196-1 REVISION 13 - DECEMBER 2010 REGULATORY GUIDE 1.196 CONTROL ROOM HABITABIL ITY AT LIGHT-WATER NUCLEAR POWER REACTORS The Licensee complies wi th the requirements in R evision 0 of this regulatory guide with the follow ing exceptions and clarifications:
The Control Room Envelope Habita bility Program is governed by Technical Specification (TS) 5.5.18, "Co ntrol Room Envelope Habitability Program," a pproved via TS Amendment No.
146 for Braidwood Station, Units 1 and 2 and TS Am endment No. 151 for Byron Station, Units 1 and 2. This TS Amen dment modified TS re quirements related to control room envelope habitability in accord ance with TS Task Force (TSTF) Traveler TSTF-448, Revision 3, "Control Room Habitability."
The implementation of a Control Room Envelop e (CRE) Habitability Program is the resul t of a regulatory commitme nt made in response to NRC Generic Letter (GL) 2003-01, "Control Room H abitability."
The CRE Habitability Program w as implemented as a re sult of findings at facilities that existing Technical Specifications may not be adequate
to ensure the requirements of 10 CFR 50 Appendix A GDC 19 are met as described in GL 2003-01.
Survey of chemical sou rces is to be performe d at least once per 6 years as part of the periodic assess ment of CRE habita bility required by TS 5.5.18.
Regulatory Guide (RG) 1.196 refe rences RG 1.78, "Evaluating the Habitability of a Nuclear Power Plant Control Ro om During a Postulated Hazardous Chemical Release," Rev ision 1. As d escribed in this Appendix to the UFSAR, Braid wood and Byron comply with Revis ion 0 of RG 1.78. Compliance with RG 1.78 is further describ ed in Section 2.2 and subsection 6.4.1.
As allowed by paragraph C.4 of Regulatory Guide (RG) 1.78, Revision 0, "The toxicity limits s hould be taken from ap propriate authoritative sources." NUREG/CR-6624 is cons idered an approp riate authoritative resource and, therefore, the toxicity limits con tained within may be used for periodic toxic gas surveys in place of those contained in RG 1.78, Revision 0.
B/B-UFSAR A1.197-1 REVISION 12 - DECEMBER 2008 REGULATORY GUIDE 1.197 DEMONSTRATING CONTROL ROOM E NVELOPE INTEGRITY AT NUCLEAR POWER REACTORS The Licensee has only co mmitted to the t esting methods a nd frequencies as specified in Sections C.1 and C.2 of Revision 0 to this regulatory guide. The requirements for determining the u nfiltered air inleakage past the CRE boundary into t he CRE is defined in Technical Specification 5.5.18 , "Control Room En velope Habitabil ity Program."
B/B-UFSAR A8.2-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 8.2 GUIDE FOR ADMINISTRATIVE PRACTICES IN RADIATION MONITORING Administrative procedures and practices of r adiation monitoring are based on 10 CFR 20 and Regulatory Guide 8.2, Revision 0.
B/B-UFSAR A8.7-1 REVISION 13
- DECEMBER 2010 REGULATORY GUIDE 8.7 INSTRUCTIONS FOR RECOR DING AND REPORTING OCCUPATIONAL RADIATI ON EXPOSURE DATA
The occupational radiation exposure record sys tem is based on Regulatory Guide 8.7, Revision 2.
B/B-UFSAR A8.8-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 8.8 INFORMATION RELEVANT TO ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AT N UCLEAR POWER STATIONS WILL BE AS LOW AS IS REASONABLY ACHIEVABLE
Regulatory Guide 8.8 describes information that is relevant to meeting the criterion th at exposures of stat ion personnel to radiation during routine operations of the station will be as low is reasonably achievable (ALARA).
Maintaining occupational radiation doses ALARA is a function of the health physics pro gram (12.5), station design (12.3), and administrative policies (12.1.1.3). The Licen see has also used operating experience during the design phase s and is utilizing supervisory personnel who have several y ears of operating experience working in its licensed stations.
The health physics p rogram includes the radiation protection program, training, and instruction. The adm inistrative policies maintain occupational exposure ALARA and establish station organization, responsibi lities, and procedures.
The station design includes access control, shielding, facility design, equipment design, airborne c ontrol, crud control, radiation monitoring waste treat ment, and modifications (based on operating experience).
The guidance provided by Regulatory Guide 8.8 (a ll issues) and by WCAP-8872, "Design, Inspection, Operatio n, and Maintenance Aspects of the Westi nghouse NSSS to Main tain Occupational Radiation Exposures as Low As Reasonably Achievable," has been used as an aid f or the radiation pro tection design.
Regulatory Guide 8.8, Re vision 3, Sections C.1, C.3, and C.4 are used as a basis for developing the ALARA and radiation protection programs with the following exce ptions: C1B p age 8.8 qualifications for radiation protection manager (RPM) job. The station does not commit to requiring the RPM to take any type of certification exam.
B/B-UFSAR A8.9-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 8.9 ACCEPTABLE CONCEPTS, MODELS, EQUATIONS, AND ASSUMPTIONS FOR A BIOASSAY PROGRAM The bioassay program w ill be in compliance with Revision 1 of Regulatory Guide 8.9, "Acceptable Concepts, Mo dels, Equations, and Assumptions for a Bi oassay Program." Bioa ssay services are
performed by either a contracted vendor or by pe rsonnel. When a vendor is contracted to perform bioassay, a requirement of the contract is in compliance with the appropria te regulatory requirements. Exelon Generation Company's b ioassay program is in compliance with the applicable regulatory requirements.
B/B-UFSAR A8.10-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 8.10 OPERATING PHILOSOPHY FOR MAINTAINING OCCUPATIONAL RADIATI ON EXPOSURES AS LOW AS IS REASONABLY ACHIEVABLE
The operating philosophy for mai ntaining occupat ional exposures ALARA is based on Revision 1 of Regulatory Guide 8.10 to the degree considered re asonable by the re spective stations.
B/B-UFSAR A8.12-1 REGULATORY GUIDE 8.12 Revision 0, December 1974
CRITICALITY ACCIDENT ALARM SYSTEMS
Area monitors are pr ovided near the spent fuel storage pool to alarm locally and in the main control room. T hese monitors will respond to radi ation in the event of a criticality accident in the new fuel storage area. These mo nitors meet the intent of Regulatory Guide 8.12 and are further described in Subsection 11.5.2.2.6.
B/B-UFSAR A8.14-1 REVISION 13
- DECEMBER 2010 REGULATORY GUIDE 8.14 PERSONNEL NEUTRON DOSIMETERS
The Nuclear Regulatory Commission has withdrawn Regu latory Guide 8.14, "Personnel Neutron Dosimeters." Revision 1 of Regulatory Guide 8.14, published in August 1977, endors ed ANSI N319-1976, "American National S tandard for Personnel Neutron Dosimeters (Neutron Energies Less Than 20 M eV)," which has been replaced by ANSI N13.52-1999, "Personnel Neu tron Dosimeters." Regulatory Guide 8.14 does not ne ed to be revised because regulations are in place that require l icensees to have an adequate dosimetry program.
Licensees are required by 10 CFR20.1501 to use dosimetry processors accredited through the National Voluntary Laboratory Accreditation Program (N VLAP). NVLAP requires p rocessors to use new standards for personnel dosi metry, ANSI N13.52-1999 and ANSI N13.11-1993, "Person nel Dosimetry Perf ormance-Criteria for Testing," to maintain an appro priate quality for dosimetry processing.
B/B-UFSAR A8.15-1 REVISION 8 - DECEMBER 2000 REGULATORY GUIDE 8.15 ACCEPTABLE PROGRAMS FOR RESPIRATORY PROTECTION
Due to changes in 10 C FR 20, Revision 0 of Regulatory Guide 8.15 is no longer applicable. Regulatory Guide 8.15, Revision 1, exceeds both the requirement s of 10 CFR 20 and the recommendations of Reg ulatory Guide 8.15, Re vision 0. The respiratory protection program will be maintained in accordance with 10 CFR 20 and Regulatory Guide 8.15, Re vision 1, except in the following areas where Guide recommendations exceed those in Revision 0:
- 1. Written procedures do not specify traini ng and minimum qualifications of re spirator program supervisors and implementation personnel. (
Reference:
Section 3.2)
- 2. Where 10 CFR 20 does not req uire internal dose monitoring, respirator users will not have internal exposu re or internal dose documented, since it is not required by 10 CFR 20. (
Reference:
Section 3.3.4)
- 3. The respirator progr am will not document res ponsibilities of each person in the p rogram, minimum trai ning and retraining requirements, or minim um qualification. (
Reference:
Section 3.5)
- 4. Inspection frequ encies for respirato ry equipment will be determined in accordance wit h applicable regulations.
Inspection frequenci es will be documented in station procedures. (Referen ce: Section 4.3)
- 5. SCBA cylinders w ill be tested and mark ed in accordance with any applicable regulat ions. (
Reference:
Section 4.3)
- 6. Respirator cartr idges that are re-used will be tested before re-use in accordance with any applicable regulations. (
Reference:
Section 4.9)
- 7. Medical evaluations for respirator use are determined in accordance with a program established by a Company physician. This program is developed to comply with applicable regulator y requirements. (
Reference:
Section 5.1)
B/B-UFSAR A8.15-1a REVISION 8 - DECEMBER 2000 8. The respirator train ing program does n ot involve "hands-on" training for all respirator types.
Training is conducted to meet applicable regula tory requirements. (
Reference:
Section 5.2)
- 9. Expectations are set for re-te sting for a resp irator fit test when an individual has a significant w eight change, but not at the point when a weight change is e xactly 10% or more. (
Reference:
Section 5.3.5)
- 10. Standby rescue perso ns are not provided for workers wearing supplied air hoods. (
Reference:
Section 6.1)
- 11. Testing frequencies and test methods for breathing air quality will be conducted in accorda nce with applicable regulations. Inspection frequen cies will be documented in station procedures. (
Reference:
Section 6.5)
- 12. Wipe samples will be taken at air connection points at the discretion of the licensee as determined necessary. (
Reference:
Section 6.5.6)
- 13. Specific time limits have no t been set regar ding length of time individuals are req uired to work while using respirators. (Refere nce: Section 6.7)
B/B-UFSAR A8.19-1 REVISION 5 - DECEMBER 1994 REGULATORY GUIDE 8.19 OCCUPATIONAL RADIATI ON DOSE ASSESSMENT IN LIGHT-WATER REACTOR POWER PLANTS DESIGN STAGE MAN-REM ESTIMATES
The dose assessment obje ctives in Revision 0 of Regulatory Guide 8.19 have been inclu ded in the UFSAR as indicated below.
Revision 0 was the cur rent revision of the regulatory guide when the operating licens e application was docketed.
Item C.(1)
The occupational radiation e xposure estimates are in Subsection 12.4.4.
Item C.(2)
B/B's radiation exposu re assessment bases are described in Subsections 12.5.3, 12.1.2.7, and 12.4.
Item C.(3)
Design changes which have resulted f rom Commonwealth Edison's dose assessment pro cess are included in Subsections 12.1.2.3, 12.1.2.7, and 12.4.
B/B-UFSAR A8.25-1 REVISION 14
- DECEMBER 2012 REGULATORY GUIDE 8.25 CALIBRATION AND ERROR LI MITS OF AIR SAMPLING INSTRUMENTS FOR TOTAL VOLUME OF AIR SAMPLED
Revision 1 of Regulatory Guide 8
.25 does not apply to nuclear power plants. The a ir sample calibration program requirements are described in UFSAR section 12.3.4.2.